The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)
The first (pilot) nuclear power plant A1 in the Slovak Republic, situated on Jaslovske Bohunice site (60 km from Bratislava) with the capacity of 143 MWel, was commissioned in 1972 and was running with interruptions till 1977. A KS 150 reactor (HWGCR) with natural uranium as fuel, D2O as moderator and gaseous CO2 as coolant was installed in the A1 plant. Outlet steam from primary reactor coolant system with the temperature of 410 C was led to 6 modules of steam generators and from there to turbine generators. Refueling was carried out on-line at plant full power. The first serious incident associated with refueling occurred in 1976 when a locking mechanism at a fuel assembly failed. The core was not damaged during that incident and following a reconstruction of the damaged technology channel, the plant continued in operation. However, serious problems were occurring with the integrity of steam generators (CO2 gas on primary side, water and steam on secondary side) when the plant had to be shut down frequently due to failures and subsequent repairs. The second serious accident occurred in 1977 when a fuel assembly was overheated with a subsequent release of D2O into gas cooling circuit due to a human failure in the course of replacement of a fuel assembly. Subsequent rapid increase in humidity of the primary system resulted in damages of fuel elements in the core and the primary system was contaminated by fission products. In-reactor structures had been damaged, too. Activity had penetrated also into certain parts of the secondary system via leaking steam generators. Radiation situation in the course of both events on the plant site and around it had been below the level of limits specified. Based on a technical and economical justification of the demanding character of equipment repairs for the restoration of plant operation, and also due to a decision made not to continue with further construction of gas cooled reactors in Czechoslovakia, a decision was made in
In this paper authors present facts about construction, operation and reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia. There was the reactor KS 150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. Two persons of personal died by suffocation with carbon dioxide. This accident has been not evaluated according to the INES scale up to present time. The second serious accident in A-1 NPP occurred in February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described in this paper. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier canal and Dudvah River as result of flooding during the decommissioning are presented. (authors)
The A-1 NPP was connected to the grid on 25 December 1972. The first serious refuelling accident happened on 5 April 1976. The second accident occurred on 22 February 1977 when a fouled fuel assembly was overheated, melted, and consequently radioactivity entered the whole primary circuit of steam generators, and a portion of it also leaked into the secondary circuit. The accident was classified as INES 4 level. There were no leaks of radioactivity into environment during the accident. A programme and schedule of bringing the A-1 NPP into a radiation-safe status was precise by subsequent decrees of the Slovak Government in 1993-1994. In January 1995, the Project was submitted to Nuclear Regulatory Authority of the Slovak Republic (NRA SR) and updated schedule of works was approved by the Government of the Slovak Republic. The radiation-safe status is characterised by minimizing of excluding a risk of negative environmental impacts of a NPP. There were 439 of total of 571 spent fuel assemblies transported to the former Soviet Union in 1984 through 1990. The remaining 132 assemblies are 'difficult-to-handle', ones the assemblies can not be taken out of their cases. Current solution the activities associated with transport and improvement of a safe storage of the spent fuel, issues related to processing, treatment, and storage of liquid radioactive wastes is represented by Bohunice treatment centre that has been built since 1993. In addition to Slovak partners, there are also foreign ones who are involved in bringing A-1 NPP into a radiation-safe status. The largest share of works is taken by British AEA Technology together with SGN, France. Other European countries are involved in A-1 NPP decommissioning, too, by providing technical assistance within PHARE programmes. Japan also provides funding of a broad programme of information exchange about NPPs decommissioning. Activities of nuclear safety inspectors were performed according to NRA SR yearly plan of inspection
In this leaflet the principle of the bituminization plant for radioactive concentrate (the intermediate liquid radioactive waste generated during the NPP A1, V-1, V-2 operations) solidification used in the Bohunice Radwaste Treatment Centre (BSC RAO) is presented
The construction history is outlined of A-1 at Jaslovske Bohunice, the first Czechoslovak nuclear power plant. Characteristic operating data of the plant in 1972-1977 are presented. The period following the second accident at the plant and its final shutdown is described. The adverse environmental impacts of the plant are characterized. A safe decommissioning of the plant is a problem that remains to be solved. (J.B.)
The equipment is described of measuring the thickness of caisson pipes built in the Bohunice A-1 reactor. The pulse-type ultrasonic thickness gauge is based on the reflection method using the double probe. The measurement accuracy is 0.1 mm. (J.B.)
The EOOS (Equipment Out Of Service) Monitor is part of the Risk and Reliability Workstation software package developed by EPRI. The software package was provided to the Jaslovske Bohunice NPP and the Slovak Nuclear Regulatory Authority within a contract with the US Department of Energy (DOE). A risk monitor for unit 3 of the Jaslovske Bohunice V2 NPP was developed by integration of a PSA model into the EOOS monitor. The paper describes the monitor and its application to risk monitoring during full power operation and reactor shutdown. (author)
The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)
In this leaflet the production of electricity in nuclear power plants, philosophy of nuclear safety with reactor WWER, influence of ionization radiation on the man, improvement on the reactor, reconstructed system on the Bohunice V-1 reactors, nuclear reactor WWER, nuclear fuel and fission reaction, are described. A briefly history of Bohunice V-1 NPP is presented
During the reconstruction of NPP V 1 Bohunice units with V-230 reactors, all in-core temperature measuring systems were upgraded. All measuring channels including thermocouples up to data acquisition system were replaced. The report provides the objectives, process and results of the reconstruction. Accuracy and reliability of the in-core measurements were enhanced, and their resistance to accident environment was achieved. Moreover metrological assurance and automatic self-check of accuracy with the temperature etalon of reactor measurements were achieved. The obtained quality of the in-core temperature measurements exceeds that of similar systems in V-213 reactors. (Authors)
This paper deals with the main activities and results of the Nuclear Power Plants Research Institute (VUJE) Trnava in the fields of development of remotely operated manipulators and robots for decontamination and dismantling. D and D of the Active Water Purifying Station (AWPS) of A-1 NPP Jaslovske Bohunice was chosen as a pilot project for the application of advanced CA technologies and manipulators for D and D tasks. The presence of radioactive, toxic or hazardous materials limits personnel access to facilities. Very often there are not enough up-to-date drawings of the installed technology. Therefore, in preparation phase of decontamination, a 3D Laser scanner and software 3Dipsos were involved as modelling technology and civil construction of the facility. Examples of acquired data and created 3D models are presented. Many D and D tasks have to be performed remotely. This paper describes the main features of developed remotely controlled manipulators. A movable manipulator MT-15 is dedicated for recognition and analysing tasks in hostile environment. A general purpose manipulator MT-80 is used for heavy duties in D and D. A long reach manipulator DENAR-41 was developed for the decontamination of underground waste storage tanks. Mock up tests of the afore-mentioned manipulators were performed before they were used in D and D tasks. Moreover the software EUCLID and IGRIP are used for simulation, analysing and optimisation of decontamination or dismantling tasks. This procedure leads to safe and more effective realisation of decontamination and dismantling tasks. The obtained results are also used for future development of suitable manipulators. The description of the initial and present state of contamination and radiation level in AWPS is presented in this paper. Experience with utilisation of advanced CA technology for acquiring as built models, development of manipulators and simulation of D and D tasks are described. (author)
The paper discuss some aspects of the main primary components lifetime evaluation program in Bohunice NPP which is performed by Nuclear Power Plant Research Institute (NPPRI) Trnava in cooperation with Bohunice and other organizations involved. Facts presented here are based on the NPPRI research report which is regularly issued after each reactor fuel campaign under conditions of project resulted from the contract between NPPRI and Bohunice NPP. For the calculations, there has been used some computer codes adapted (or made) by NPPRI and the results are just the conclusive and very brief, presented here in Tables (Figures). (authors)
Replacement of thermocouples in the protection tube blocks was a key phase of the reconstruction of in-reactor temperature measurements at Bohunice V-1 with regard to the success, reliability and impact on safety of unit operation. The replacement consisted of reliable and safe withdrawal of 216 old thermocouples, their disposal and installation of new thermocouples into dry channels. In the material presented, this phase of reconstruction is described in details, with focus on the evaluation of replacement quality and check activities carried out at the new installed thermocouples. (Authors)
In this leaflet the Bohunice Radwaste Treatment Centre (BSC RAO) is presented. BSC RAO is designed to process and treat liquid and solid radwaste, arising from the NPP A-1 decommissioning, from NPPs V-1, V-2, and Mochovce operations, as well as institutional radwaste of diverse institutional (hospitals, research institutes) in the Slovak Republic. Transport, sorting, incineration, compacting, concentration and cementation of radwaste as well as monitoring of emission are described
The Siemens in-core sipping system has proved to work satisfactorily in the Jaslovske Bohunice Nuclear Power Station. In 1990, Siemens (KWU) installed a new version of the system advanced in the light of past operating experience, in which the rectangular eightfold bell had been replaced by rotationally symmetrical sevenfold bell. The number of failed fuel elements detected in the four generating units of the Jaslovske Bohunice Nuclear Power Station is relatively small, documenting reliable operation of the fuel elements in the WWER-440 reactor. (orig./HP)
The automated system of maintenance control at the V-1 nuclear power plant at Jaslovske Bohunice uses experience gained with the maintenance of the A-1 nuclear power plant. With regard to the range of work operations, maintenance includes inspection, routine repair, overhaul of equipment and replacement. Also observed is the classification of equipment according to whether it may be repaired without reactor shutdown or whether the reactor will have to be shut down. At present the maintenance of the Jaslovske Bohunice nuclear power plant is being processed by an automated control system into five year variable plans of repairs, annual and monthly plans of repairs, plans of shut-downs and a schedule of unit shutdowns. The repair plan includes over 6000 items. (Z.M.)
Long term operating experience of Bohunice units maintenance activities are overviewed in the paper. Based on common experience of WWER NPP operators, separate maintenance department was established at Bohunice NPP in very early stage of plant operation. Maintenance management, maintenance planning, outage management, diagnostics and monitoring, inspection technologies and backfitting activities are described particularly to demonstrate the capability of Bohunice maintenance department for most complex repairs and maintenance works of nuclear power plant components and equipment, including reactor and turbine itself. (author)
The upgrading and safety enhancement of both the Bohunice V-1 and V-2 reactors is described in detail. The total estimated cost of the gradual reconstruction of these two units during 1996 to 1999 is 180 mil. US dollars. For the 1995 to 1997 period, the actions common for both units include a quality assurance programme, a personnel training programme, installation of a multifunction simulator, implementation of symptom-oriented operation procedures, installation of diagnostic systems, of a site security system, and of a teledosimetric system. At present, the main maintenance tasks are: to carry out major repair of units, to remedy service interruptions, to enhance equipment service availability, to enhance the technical level of corrective actions at equipment. Investment into maintenance level upgrade has grown from 7.5 mil. Slovak crowns in 1994 to estimated 32 mil. in 2000. The partners of international cooperation are mentioned. (M.D.)
Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m2. It follows that the total gamma contamination is of the order of 1014 to 1015 Bq and total alpha contamination 1011 to 1013 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)
Slovakia remains significantly dependent on imports of primary energy sources, which represent as much as 80% of the demand. Of the total consumption of electricity in Slovakia, 40% was generated in nuclear power plant units in 1998. Slovakia operates 6 units with WWER 440 nuclear reactors. Slovakia is the signatory of all important international agreements and conventions in the field of nuclear energy, and its legislation is in an advanced stage of approximation to European Union law. This is a very important aspect, showing Slovakia's approach to nuclear safety. In 1993 Slovakia accepted the commitments of the UN Convention on Climate Changes, including a reduction of greenhouse gases to 1990 levels by the year 2000. Moreover, as an internal target Slovakia has set the reaching of the 'Toronto Objective', i.e. 20% reduction in COx, emissions through the year 2005 as compared to 1988. In our opinion, this is not possible without nuclear energy. Time has shown, that the political aspects are more powerful, especially if you underestimate their importance over the than the technical ones. In the case of Bohunice V-1 NPP the political aspects were on the following levels: 1. Slovak republic (Czechoslovakia), political changes, decisions of the government; 2. European Union - Agenda 2000, Accession criteria, nuclear safety criteria, EBRD; 3. Austria as a neighbouring country. Starting with year 1990, 23 expert missions took place at Bohunice V-1 NPP by now. The only criteria for further operation should have been Nuclear safety, which is supervised by NRA SR. It was fully in compliance with EU policy, each country is solely responsible for its energy sector and for nuclear energy use. Our satisfaction lasted not too long. Following negotiation with EU on the highest political level, driven by willingness to be invited for negotiation of accession on the Helsinki Summit, the Slovak government decided on September 14th, on Bohunice V-1 Units shutdown in 2006 and 2008
Communication is the base of everyday existence of a modern person and every company. It is not easy to work in this area in a changing 'eastern' country. Many tools, which are used are in the mind of people connected with 'propaganda'. I would like to share our experience with you. The goal of an internal communication is to spread and provide continuous current of objective information between the management of Bohunice NPPs and its personnel and between the personnel itself. Communication with the Bohunice NPPs employees helps to get acquainted with their opinions and ideas concerning the subsidiary and nuclear power industry
A brief overview is given of the power sources in Slovakia which include 6 operational reactor units (4 at Jaslovske Bohunice and 2 at Mochovce) and 2 units under construction (at Mochovce). The efforts undertaken in the past 10 years and aimed at upgrading the nuclear safety of the two older (V-230 Soviet type) units at Bohunice are highlighted. The relevant regulatory decisions are dealt with and the measures already carried out are listed. Also characterized are several IAEA international safety assessment missions and safety-aimed meetings which took place in 1998 and 1999 and are of concern to the older Bohunice units. (A.K.)
It is not generally known that the first serious failure of nuclear power plant (NPP) technology with loss of human lives occurred in NPP Jaslovske Bohunice (Czechoslovakia) in January 1976. A year later the second accident finally broken reactor A1 with large radioactive contamination. This material was later (in 1980) washed into the nearby drainage by the heavy rain. In cleaning procedure, the contaminated soil particles contaminated the slopes of the drainage. These spots have the shape of 'blurs' about 15 cm wide with a scale of contamination from 0,067; 0,15; 2,38; 9,5; 45.5 up to 322 kBq/kg 137Cs. The research was done in cooperation with the Institute of Tumorbiology, University of Vienna, within the grant Action Austria - Slovak Republic. Details of radioactivity at the area were obtained thanks to the Research Institute of the Nuclear Energy in Trnava, Slovakia. In our ten years long-term study of contaminated soil around nuclear power plant (NPP) Jaslovske Bohunice 24 species of local flora were used to show impact of these accidents. The 19 km long banks of the Jaslovske Bohunice NPP waste water recipient has been identified as contaminated by 137Cs. In total, more than 67,000 m2 of river banks have been found as being contaminated at levels exceeding 1 Bq 137Cs/g of soil. Used phytotoxic and cytogenetic -in situ' tests were extended by analyses of pollen grains. Although the dose of some samples of radioactive soil was relatively high (322 kBq kg-1) no any significant impact on the biological level of tested wild plant species was observed. Possible explanation (such as adaptation and resistance) is discussed. (author)
Bohunice nuclear Power Plant generation represents almost 50% of the Slovak republic electric power production. Due to such high level of commitment to nuclear power in the power generation system, a special attention is given to safe and reliable operation of NPPs. Safety upgrading and operational reliability improvement of Bohunice V-1 NPP was carried out by the Bohunice staff continuously since the plant commissioning. In the 1990 - 1993 period extensive projects were realised. As a result of 'Small Reconstruction of the Bohunice V-1 NPP', the standards of both the nuclear safety and operational reliability have been significantly improved. The implementation of another modifications that will take place gradually during extended refuelling outages and overhauls in the course of 1996 through 1999, is referred to as the Gradual Reconstruction of the Bohunice V-1 Plant. The general goal of the V-1 NPP safety upgrading is the achievement of internationally acceptable level of nuclear safety. Extensive and financially demanding modification process of Bohunice V-2 NPP is likely to be implemented after a completion of the Gradual Reconstruction of the Bohunice V-1 NPP, since the year 1999. With this in mind, a first draft of the strategy of the Bohunice V-2 NPP upgrading program based on Probabilistic Safety assessment consideration was developed. A number of actions with a general effect on Bohunice site safety is evident. All these activities are aimed at reaching the essential objective of Bohunice NPP Management - to ensure a safe, reliable and effective electric energy and heat generation at the Bohunice site. (author)
The early closure of the Bohunice V1 nuclear power plant following a time schedule that the Slovak government agreed with the EU three years ago is a political obligation of the government. The only objections to this plan so far have been related to the weakening of the generation base and economy of Slovenske elektrarne (SE), a.s. Neither the Nuclear Regulatory Authority of the Slovak Republic, nor the International Atomic Energy Agency, nor foreign experts preparing the closure of V1 in Jaslovske Bohunice have raised objections relating to increased security risks. Since he took up office last summer, the Minister of Economy Pavol Rusko has not hidden his personal interest in revising the agreement and extending the operation of the nuclear power plant. The fact that closing the power plant would have a negative impact on Slovakia's economy is not an argument good enough to persuade the original EU members. A proven security risk would be a better argument, especially as the operation of the V1 plant would not be extended literally - rather one of the reactors would work two years longer. Risk increased by 100%, a risk close to the level of acceptability in the EU - that is how in October he described the risks related to closure of V1 units according to the agreed time schedule, i.e. the first unit to be shut down by December 31, 2006 at the latest and the second unit by December 2008. The Minister based his opinion on a study prepared by the engineering and consulting company - Relko that assessed the risks V1 would represent after the first unit has been shut down. 'The only solution is to shut down both units at the same time', he concluded and in a comment addressed to the EU he added: 'If the EU has sincere intentions towards Slovakia, respects its security and does not wish it economic problems it will accept this'. (author)
This CD is multimedia presentation of programme safety upgrading of Bohunice V1 NPP. It consist of next chapters: (1) Introductory speeches; (2) Nuclear power plant WWER 440; (3) Safety improvement; (4) Bohunice Nuclear power plants subsidiary; (5) Siemens; (6) REKON; (7) VUJE Trnava, Inc. - Engineering, Design and Research Organisation; (8) Album
Slavik, O.; Moravek, J.; Stubna, M.
For assessments of hard-to-detect radionuclides (HD-RN) contents in various type of radwastes at the NPP-A1, available empirical data referenced to 137Cs (actinides, 90Sr, 99Tc, 63Ni, 14C) and the theoretical assessment for the remaining HD-RN using calculated RN inventory and a simple model with effective relative (137Cs) spent fuel release fractions was applied. The analytical data of extended radiochemical analysis for the existing available operational radwaste forms have been reviewed for this purpose. 137Cs, 90Sr and 241Am were set up as release markers for partial spent fuel release groups of HD-RNs within which the total fractions of HD-RN released to the operational radwastes were assumed to be constant. It was shown by the assessment carried out that 137Cs and HD-RNs 129I, 99Tc, and partly 79Se and 14C are the main contributors to the disposal dose limit for the radioactive concentrate at NPP A-1. In the case of the radioactive sludge from the operational radwaste system the role of predominant dose contributors belongs to actinides 239,240Pu and 241Am. In the case of clearance of radioactive material from the NPP-A1 site, only the reference radionuclide, 137Cs was predicted to be the most dominant dose contributor. In all of these cases the estimated contributions of other hard-to-detect radionuclides to respective disposal or release dose limit are lower by 2 and more orders of magnitude. As a lesson learned, the most attention is proposed to focus on the control and measurement of the critical HD-RNs indicated by the assessment. For the control of less important HD-RNs, the developed release coefficient method is sufficient to be applied.
Paper summarizes experience from last 15 years of operation at NPP Jaslovske Bohunice. During this period, leaking fuel assemblies have had been identified by in-core sipping method and verified by vendor specified canister sipping method. Methodology of operational and outage fuel integrity monitoring is described. Full survey of identified leaking assemblies is given. Fuel failure rates are calculated separately for V-1 (V-230 type) and V-2 (V-213 type) units. Systematic difference - significantly lower fuel failure rate at V-213 units exists for all period investigated. Analysis of potential fuel failure reasons and all related measures (planned and already implemented) are presented. Design, operation and fabrication features have been analyzed with the aim to identify dominant factors contributing to fuel failure. No unambiguous reasons have been found so far. It is believed that there is a superposition of several factors and differences causing higher failure rate at V-230 type units. (author)
The sequences in development of computer equipment s and the sequences in development of measurement tools and procedures are listed in submitted presentation - from start-up the power plant in operation until present days. Present status of integration of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors in NPP V1 is presented here. The ways of data acquisition, storing of results and their evaluation are described in this presentation. In conclusion some practical possibilities of using a a monitoring system for accuracy and reliability are listed. (Authors)
In this paper and in presentation some results of upgrading of the NPP Bohunice V-1 are presented. For the first time, extensive upgrades are performed in all safety-related areas of both units with VVER 440/230 reactors. These upgrades focused on: - Expansion and upgrading of the process safety systems; - Replacement of the safety I and C system with a TELEPERM XS-based system; - Spatial separation of safety equipment; - Modernisation of the electrical auxiliary power systems; - Seismic upgrading and fire protection; - Improvement of the man-machine interface. This upgrade is considered exemplary around the world. The most extensive stage of gradual reconstruction of Unit 2 was completed according to the schedule in January 1999. For the first time, a reactor which incorporates state-of-the-art digital I and C in its reactor protection system is on-line. (author)
Bohunice V1 in Slovakia is a Russian designed two unit WWER 440, Model 230 Pressurized Water Reactor. The plant was not originally designed for earthquake. Subsequent and ongoing reassessments now confirm that the seismic hazard at the site is significant. EBO, the plant owner has contracted with a consortium lead by Siemens AG (REKON) to do major reconstruction of the plant to significantly enhance its safety systems by the addition of new systems and the upgrading of existing systems. As part of the reconstruction, a complete seismic assessment and upgrading is required for existing safety relevant structures, systems and components. It is not practical to conduct this reassessment and upgrading using criteria applied to new design of nuclear power plants. Alternate criteria may be used to achieve adequate safety goals. Utilities in the U.S. have faced several seismic issues with operating NPPs and to resolve these issues, alternate criteria have been developed which are much more cost effective than use of criteria for new design. These alternate criteria incorporate the knowledge obtained from investigation of the performance of equipment in major earthquakes and include provisions for structures and passive equipment to deform beyond the yield point, yet still provide their essential function. IAEA has incorporated features of these alternate criteria into draft Technical Guidelines for application to Bohunice V1 and V2. REKON has developed plant specific criteria and procedures for the Bohunice V1 reconstruction that incorporate major features of the U.S. developed alternate criteria, comply to local codes and which envelop the draft IAEA Technical Guidelines. Included in these criteria and procedures are comprehensive walkdown screening criteria for equipment, piping, HVAC and cable raceways, analytical criteria which include inelastic energy absorption factors defined on an element basis and testing criteria which include specific guidance on interpretation
In this leaflet results of exploitation of four units of the Bohunice V-1 and V-2 NPPs are presented. The electricity and heat production in July 2004 are reviewed. Within a July 2004 the electricity was produced: 285 GWh (block 1), 292 GWh (block 2), 0 GWh (block 3), 20 GWh (block 4), totally 597 GWh, and 6352 GWh within a January - July 2004. The heat production in July 2004 was 9 417 GJ, and within a January - July 2004 it was produced 941 403 GJ of heat. After enlargement of European Union (EU) by ten new member states on May 1, 2004 the number of nuclear units has been risen by 19 units. 136 nuclear units were in operation in 'old' European Union. The most of nuclear units have been brought by Czech Republic and Slovak Republic (6 operational units, each state); Hungary has brought four units, Lithuania two and Slovenia one nuclear unit. Remaining five countries - Poland, Estonia, Cyprus, Latvia and Malta do not use nuclear energy for electricity production. Nuclear energetics is used in fourteen countries in enlarged European Union. France operates the largest number of nuclear units (59), which generated 77.67 per cent electricity of total French produced electricity in last year. However, France has lasted its dominant position in European Union since May, because two Lithuanian nuclear units generated 79.88 per cent of electricity in last year. In 2003 nuclear units reached 33.1 per cent of electricity generation within European Union. After enlargement of EU in 2004 this share should be raised to 34 per cent
Of the total consumption of electricity in Slovakia, 42% was generated in nuclear power plant units in 1999. Slovakia operates 6 units with a WWER 440 nuclear reactors, 4 of them are at Bohunice site and 2 at Mochovce. The Nuclear Regulatory Authority of SR is not the only regulatory body controlling nuclear activity. Both - the system of nuclear activities regulation in Slovakia as well as the approach to Nuclear Safety enhancement of the operator were positively judged by IAEA and WENRA. In 1993 -Slovakia has accepted the commitments of the UN Convention on Climate Changes, including a reduction of greenhouse gases to 1990 levels by the year 2000. Moreover, as an internal target Slovakia has set the reaching of the ,'Toronto Objective', i.e. 20% reduction in COx emissions through the year 2005 as compared to 1988. Taking into account the actual situation as well as natural conditions for some renewable sources utilisation, the target won't be reached without nuclear energy. The nuclear energy is free of emissions, does not burn oxygen, and with the share of production in Slovakia will remain significant contributor. To the environment protection it contributes also by replacing fossil heat plants with heat delivery for the region. In case of radiological wastes the environment protection is ensured by very strict system of control, evidence, treatment and repository. To conclude, Bohunice NPPs were, are and will remain very important part of the Slovak's economy, creating conditions for its (sustainable) development
The paper presents the findings of a study of the role of the human factor based on the results of the first four years of operation of the Bohunice B-1 nuclear power plant. It describes the method by which plant personnel are trained and the system of maintaining the level of staff skills. It is expected that there will be an improvement in the quality of personnel training and that an analysis of the role of the human factor will be made in the course of subsequent operation. (author)
The paper presents testing methodology of neutron flux measurements systems used standard chains incorporated to the reactor control system and non-standard measuring system, used during start-up tests after refuelling of the reactors. Comparison of both measuring systems is given. Methodology is illustrated on results of measurements performed at Bohunice NPP during reactor start-up tests.(author)
Procedure for education and training of all the personnel employed at Bohunice Nuclear power plant is presented in detail describing the training system structure, kinds of training, staff members qualification development, short term and long term tasks needed to assure attaining the training objectives. The proposed Staff Members Lifetime education implementation project contains basic starting points, measures to be implemented by 1998. It was prepared on the basis of a primary analysis which confirmed the existing need for implementing the lifetime education system
Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)
The methods are compared of electroslag welding and of arc welding with a view to their possible application in welding the Bohunice A-1 reactor pressure vessel. Considered are the thermal deformation effects of welding on the physical properties and the stress present in welded joints. For testing, plates were used having the dimensions of 1100x2300x200 mm and rings with 4820 mm outer diameter, 1800 mm height and 170 mm thickness made of steel CSN 413O30 modified with Ni, Al+Ti. The deformation effect of welding on the residual surface and triaxial stress, the specific stored energy, the initiation temperature of brittle crack and the critical size of the initiation defect corresponding to the thermal deformation effect of welding were determined. It was found that for electroslag welding, there is a low probability of crack formation in the joints, a low level of residual stress and a low level of specific stored energy in a relatively wide joint zone. For arc welding there is a considerable probability of defect formation in the vicinity of the sharp boundary of the joint, a high level of the triaxial state of stress in the tensile region, and a high level of specific stored energy concentrated in the narrow zone of weld joints. The recommended thermal process is given for welding pressure vessels made of the CSN 413030 steel modified with Ni, Al+Ti, and 150 to 200 mm in thickness. (J.P.)
A seismic assessment and strengthening investigation is being performed for selected structures at the Bohunice V1 Nuclear Power Plant in Slovakia. Structures covered in this paper include the reactor building complex and the emergency generator station. The emergency generator station is emphasized in the paper as work is nearly complete while work on the reactor building complex is ongoing at this time. Seismic evaluation and strengthening work is being performed by a cooperative effort of Siemens and EQE along with local contractors. Seismic input is the interim Review Level Earthquake (horizontal peak ground acceleration of 0.3 g). The Bohunice V1 reactor building complex is a WWER 4401230 nuclear power plant that was originally built in the mid-1970s but had extensive seismic upgrades in 1991. Siemens has performed three dimensional dynamic analyses of the reactor building complex to develop seismic demand in structural elements. EQE is assessing seismic capacities of structural elements and developing strengthening schemes, where needed. Based on recent seismic response analyses for the interim Review Level Earthquake which account for soil-structure interaction in a rigorous manner, the 1991 seismic upgrade has been found to be inadequate in both member/connection strength and in providing complete load paths to the foundation. Additional strengthening is being developed. The emergency generator station was built in the 1970s and is a two-story unreinforced brick masonry (URM) shear wall building above grade with a one story reinforced concrete shear wall basement below grade. Seismic analyses and testing of the URM walls has been performed to assess the need for building strengthening. Required structural strengthening for in-plane forces consists of revised and additional vertical steel framing and connections, stiffening of horizontal roof bracing, and steel connections between the roof and supporting walls and pointing of two interior transverse URM
Based on the 1991 recommendation by the former Czechoslovak nuclear regulatory body - the Czechoslovak Atomic Energy Commission - the minor reconstruction of the V-1 nuclear power plant at Jaslovske Bohunice was aimed at safety improvement in the following fields: reactor pressure vessel and primary circuit integrity, hermetic compartments, instrumentation and control systems and accident protection systems, home consumption electrical systems, fire safety, seismic resistance, and reactor aftercooling in case of steam generator feedwater failure. The results of the reconstruction are presented. The reconstruction provided for all the recommendations. (J.B.). 2 tabs
Slugen, V.; Lipka, J.; Dekan, J.; Tóth, I.; Smieško, I.
Steam generators of four VVER-440 units at nuclear power plants V-1 and V-2 in Jaslovske Bohunice (Slovakia) were gradually changed by new original "Bohunice" design in period 1994-1998. Corrosion processes before and after these design and material changes in Bohunice secondary circuit were studied using Mössbauer spectroscopy during last 25 years. Innovations in the feed water pipeline design as well as material composition improvements were evaluated positively. Mössbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens scrapped from water pipelines or in form of filter deposits. The corrosion of new feed water pipelines system (from austenitic steel) in combination to innovated operation regimes goes dominantly to magnetite. The hematite presence is mostly on the internal surface of steam generator body and its concentration increases towards the top of the body. In the results interpretation it is necessary to consider also erosion as well as scope and type of maintenance activities. The long-term study of phase composition of corrosion products at VVER reactors is one of precondition for the safe operation over the projected NPP lifetime.
In this book the history of constructing, main characteristics, complex examination, physical commissioning, energetic commissioning, exploitation, active zone, fuel elements, nuclear reactor, primary circuits, circuits of heavy water, secondary circuits, gaseous management, transport-technologic equipment, measurement and regulation, electric schema, dosimetry, system of tightness control of covering fuel elements, equipment of chemical technology and chemical regime, emergency circuits of the reactor, serious accident in A1 NPP, leakages of vapour-generators of the A1 nuclear power plant in Jaslovske Bohunice (the Slovak Republic) are described.
Due to several deficiences of the WER Model 230 type reactor a leak before break demonstration of this reactor is of primary importance. The complex project for NPP V1 Jaslovske Bohunice includes a static and dynamic stress analysis of the primary piping, a fatique damage analysis, leak rate assessments and an analysis of the stability of the heavy components supports. The material database includes data on fracture mechanics, on assessment of corrosion properties, and on the influence of 100 000 hr service exposure on base metal and welds including disimilar welds. The program was supported by large scale experiments on RPV safe-end, pressurizer safe-end, elbow welds with through-wall cracks and leak rate measurements. The results and applications are discussed. (orig.)
As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm2 the welded joints in the reactor core are exposed to an integral dose of 3x1018 n/cm2. The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)
The paper summarizes some facts of risk-informed regulation developments within UJD regulatory environment. Based on national as well as international operating experience and indications resulted from PSA, Nuclear Regulatory Authority of the Slovak Republic (UJD) since its constituting in 1993 has devoted an effort to use PSA technology to support the regulatory policy in Slovakia. The PSA is considered a complement, not a substitute, to the deterministic approach. Suchlike integrated approach is used in decision making processes and the final decision on scope and priorities is based on it. The paper outlines risk insights used in the decision making process concerning Bohunice NPP safety upgrading and focuses on the role of PSA results in Gradual Reconstruction of Bohunice VI NPP. Besides, two other examples of the PSA results application to the decision making process are provided: the assessment of proposal of modifications to the main power supply diagram (incorporation of generator switches) and the assessment of licensee request for motor generator AOT (Allowable Outage Time) extension. As an example of improving support of Bohunice V-2 risk-informed operations, concept of AOT calculations and Bohunice V-2 Risk Monitor Project are briefly described. (author)
The approach of the Bohunice nuclear power plant to the safety culture issue is highlighted. Activities performed so far at the plant to improve the plant safety culture with a view to enhancing the awareness of each employee and thus to minimize the effect of the human factor on the evolution of incidents and accidents at the plant are described. (author)
In this presentation author describes the nuclear reactor A-1 in Jaslovske Bohunice (Slovakia). Author analyzes two reactor accidents which took off at this reactor. The first accident proceeded on January 5, 1976 during exchange of fuel elements when coolant - carbon dioxide - escaped. The second serious accident became on February 22, 1977 again during exchange of spent fuel elements. At this accident moderator - heavy water penetrated into the primary circuit of the reactor. Heavy water was subsequently removed from the reservoirs into the reserve tank in order not to leak out into the primary circuit. Inserting fuel element was melted. This accident was evaluated as grade 4 on seven-grade the international INES scale. A crash course and course parameters of the both accidents are analyzed.
The paper brings basic information on geological and seismic characteristics of the site of NPP Jaslovske Bohunice, Slovakia. Western Carpathians and Trnava, bay geological properties are briefly introduced. The most important macroseismic data and data obtained from field measurements are analysed. Main features of the expected strong seismic motion are discussed. The attention is devoted to local soil characteristics just under the site of NPP. (author)
This paper presents the results of safety analysis of a medium LOCA (break size 100 mm in cold leg) for the V2 Bohunice nuclear power plant (VVER-440/V-213), and compares the results calculated by various computer codes (MELCOR, MAAP, RELAP/SCADAP). The analysis is performed within the SWISSLOVAK project by the safety analysis group at the Nuclear Regulatory Authority of the Slovak Republic. The medium LOCA accident is combined with station blackout scenario which leads to the core uncovery and meltdown of the reactor core. The core meltdown is followed by the core relocation to the lower plenum, heat up of the reactor pressure vessel lower head, failure of the lower head, and debris ejection into the reactor cavity. The time of key events calculated by various computer codes is similar. The start of core melt is predicted within 0.8 to 1.08 hours and the reactor pressure vessel lower head failure is predicted within 4.1 to 6.3 hours since the initiation of the accident. A substantial release of noble gases to the environment through the permanent containment leakage is calculated. The compartmentalization of the containment and the presence of the bubble condenser affect the release of the fission products. (author)
For nuclear power plants which have been in operation for more than 15 years, backfitting or even complete replacement of the instrumentation and control (I and C) equipment becomes an increasingly attractive option, motivated not only by the objective to reduce the cost of I and C system maintenance and repair but also to prolong or at least to safeguard the plant life-time: optimized spare-part management through use of standard equipment; reduction of number and variety of different items of equipment by implementing functions stepwise in application software; adding new functionality in the application software which was not possible in the old technology due to lack of space; safeguarding of long-term After-Sales-Service. Some years ago Bohunice V1 NPP, Slovak Republic and Paks NPP, Hungary intended to replace parts of their Safety I and C, mainly the Reactor Trip System, the Reactor Limitation System and the Neutron Flux Excore Instrumentation and Monitoring Systems. After a Basic Engineering Phase in Bohunice V1 and a Feasibility Study in Paks both plants decided to use the Siemens' Digital Safety I and C System TELEPERM XS to modernize their plants. Both Bohunice, Unit 2 and Paks, Unit 1 have been back on line for over six months with the new Digital Safety I and C. At the present time Bohunice, Unit 1 and within the next few months Paks, Unit 2 will be replaced. Trouble-free start-ups and no major problems under operation in the first two plants were based on: thorough understanding of the VVER 440 technology; comprehensive planning together with the plant operators and authorities; the possibility to adapt TELEPERM XS to every plant type; the execution of extensive pre-operational tests. Regarding these modernization measures Siemens, as well as the above Operators, have gained considerable experience in the field of I and C Modernization in VVER 440 NPPs. Important aspects of this experience are: how to transfer the VVER technology to TELEPERM XS; how to
The contribution deals with form, present state and results of Nuclear Power Plants Research Institute (the Slovak acronym is VUJE - Vyskumny Ustav Jadrovych Elektrarni) participation in the NPP V-2 Jaslovske Bohunice Modernization and Safety Improvement Project. Short description of VUJE history, activity and results is also done as well as NPPs Jaslovske Bohunice characterization. (authors)
The 18 km long banks of the Bohunice NPP waste water recipient are contaminated by 137Cs as a result of two accidents on the CO2 cooled NPP-A1 unit in 1976 and 1977. Contamination acceptance limits 6 or 8 Bq 137Cslg of soil, depending on contaminated area size, were derived on the basis of developed principles, and approved by the authorities. Removing and safe burial of 1,100 m3 of contaminated soil from steep area and 15 cm thick clean soil covering on about 1ha of flat area of the contaminated banks is planned in frame of the re-considered restoration project implementation in 1995/96. (author)
The 18 km long banks of the Bohunice NPP waste water recipient are contaminated by cesium-137 as a result of two accidents on the CO2 cooled NPP A1 unit in 1976 and 1977. Since 1992, all he contaminated waste waters dumping from NPP Bohunice has been carried out directly to the Vah River through a specially constructed 15 km long pipeline. The final extent of contamination in the Bohunice site is represented. The overall contaminated area in this site with cesium-137 activity above 1 Bq/g of soil is about 67000 m2 and thus, the corresponding volume of top 20 cm thick soil layer is about 13000 m3. For optimizing less costly remedial measures (warning signs...) an agreed scenario with a pre-estimated factor factor collective dose 2.10-7 man.Sv.y-1/(m2.Bq137Cs.g-1) was applied. Limitation of individual effective doses according to a site specific stay scenario was also considered for this purposes with a limiting value of 0.25 mSv/y. Cost analysis of available remedial techniques were carried out, too. Two techniques have been selected for the contaminated banks restoration project: 1) removing/disposal of 20 cm soil top layer from steep and unengineered banks, and 2) mechanical dilution/fixation of contamination by clean 15 cm soil cover for the contaminated flat areas. Two-fold reduction of anticipated potential radiation risk were accepted, maximally, for the lastly mentioned technique, however cost saving is considerable (about 10-time lower the cost comparing to removing/disposal technique one). The basic acceptance limits AL for 137Cs in soil and criteria size of continuously contaminated bank areas were derived as: AL200 = 6.0 Bq/g and 800 m2 (300 m) or AL50 8.0 Bq/g and 200 m2 (80 m) for removing/disposal of the soil on steep unengineered banks. For clean soil covering technique the resulting limits are in an interval AL50C = 8 up to 16 Bq/g. According to the criteria developed, it is necessary to subject to restoration about 11000 m2 of contaminated area on
This presentation describes approach used for nuclear safety assessment at the V -1 Bohunice after the Gradual Upgrading. The deterministic and probabilistic results are presented in detail form. (author)
In this book the history of construction, commissioning and exploitation of NPP A1, NPP V1 and NPP V2 in Jaslovske Bohunice is presented on documentary photos. Vicinity around of these NPPs is presented, too
After successful completion of extensive Gradual Reconstruction (1996-2000) of Nuclear Power Plant V1 (2 x VVER440/V230), there started modernisation project of Nuclear Power Plant V2 (2 x VVER440/V213) in the site Jaslovske Bohunice with planned completion in the year 2008. The main goal and priority of NPP V2 modernisation programme is to increase safety and reliability of the operation, but also to create conditions for extension of the operating life and economy improvement of NPP V2 operation. NPP V2 units in Jaslovske Bohunice were commissioned in the year 1984 and 1985. In the year 1997, management of Slovenske Elektrarne approved goals of modernisation programme and safety increasing of NPP V2. At modernisation programme defining, there were taken into consideration results of safety assessment and recommendations for improvement of NPP with VVER 440/V213 processed within the projects IAEA, WANO, VUJE and the other organisations which have had experiences with the operation of NPP with VVER reactors (basic documents: Safety report NPP V2 after 10 years of operation, VUJE, 1993, Safety issue and their ranking for nuclear power plants WWER 440/V213 type, IAEA, 1996, Safety improvements of NPP V2 and design of their solution, VUJE, 1997). Detail range, content and schedule of programme implementation were elaborated by VUJE in the year 2001. VUJE worked out solution designs into the level of project requirements (Conceptual Design) in the document: 'Safety concept for modernisation and safety increasing of NPP V2' Modernisation works are implemented mainly in I and C and electro part, works in nuclear systems and the civil part are implemented in smaller range. Implementation works in modernisation project are realized mainly during planned units outages for refuelling. VUJE as the general designer provides elaboration of design documentation, safety documentation; support of general contractor and it is responsible for overall coordination and functionality
This paper provides a description of water chemistry monitoring and diagnostic system installed at Slovak NPP Jaslovske Bohunice. System has complex architecture and covers laboratory data, chemistry and radiochemistry on-line monitoring data, process data acquisition and processing and diagnostics. Pre-filtered data from process computer and chemistry on-line monitors are recorded together with laboratory data in the ORACLE-based information system CHEMIS with many presentation and processing features. Brief information is given about the basic features of a newly developed diagnostic system for early detection and identification of anomalies incoming in the water chemistry regime of the primary and secondary circuit of VVER-440 type unit. This system, called SACHER (System of Analysis of Chemical Regime) has been installed within the major modernization project at the NPP Bohunice in the Slovak Republic. System SACHER has been developed fully in MATLAB environment. Diagnostic system works exclusively with available on-line data as an operation personnel support application allowing effective response to adverse chemistry events/trends. The availability of prompt information about the chemical conditions of the primary and secondary circuit is very important in order to prevent the undue corrosion and deposit build-up processes within the plant systems. The typical chemical information systems that exist and work at the NPPs give the user values of the measured quantities together with their time trends and other derived values. It is then the experienced user's role to recognize the situation the monitored process is in and make the subsequent decisions and take the measures. The SACHER system, based on the computational intelligence techniques, inserts the elements of intelligence into the overall chemical information system. It has the modular structure with the following most important modules: - normality module- its aim is to recognize that the process
Full Text Available One of the main goals of the nuclear industry is to increase the nuclear safety and reliability of nuclear power plants (NPPs. As the steam generator (SG is the most corrosion sensitive component of NPPs, it is important to analyze the corrosion process and optimize its construction materials to avoid damages like corrosion cracking. For this purpose two different kinds of SGs and its feed water distributing systems from the NPP Jaslovske Bohunice were studied by nondestructive Mössbauer spectroscopy. The samples were scraped from the surface and analyzed in transmission geometry. Magnetite and hematite were found to be the main components in the corrosion layers of both SGs. Dependant of the material the SG consisted of, and the location in the system where the samples were taken, the ratios between magnetite and hematite and the paramagnetic components were different. The obtained results can be used to improve corrosion safety of the VVER-440 secondary circuit as well as to optimize its water chemistry regime.
Gonzalez Fernandez-conde, A.; Brochet, I.; Ferreira, A.
From October 2003 until december 2014 the Consortium consisting of Iberdrola Engineering and Construction (leader). Empresarios Agrupados Internacional, and Indra Sistemas has carried out the project Project Management Unit ((PMU) for the decommissioning of Bohunice V1 NPP (units 1 and 2), type VVER-440/V-230 in Slovakia. during the first phase (2003-2007) EdF was also part of the Consortium. The project is funded by the Bohunice International Decommissioning Support Fund (BIDSF) administered by the RBRD. The main objective of the project is to provide the necessary engineering and resources of project management for planning, execution, management, coordination and monitoring of all tasks in support of the decommissioning. (Author)
Described is the common project for completion of bubbler condenser qualification for nuclear power plants in Bohunice, Mochovice, Dukovany and Paks. Functionality of the bubbler condenser was elaborated during the simulation of the main steam line brake, medium break and small break LOCA. On this basis the appropriate operation of bubbler condenser containment under accident conditions can be positively confirmed
The level 2 PSA model of the J. Bohunice V1 NPP was developed in the RISK SPECTRUM Professional code with the following objectives: to identify the ways in which radioactive releases from the plant can occur following the core damage; to calculate the magnitudes and frequency of the release; to provide insights into the plant behaviour during a severe accident; to provide a framework for understanding containment failure modes; the impact of the phenomena that could occur during and following core damage and have the potential to challenge the integrity of the confinement; to support the severe accident management and development of SAMGs. The magnitudes of release categories are calculated using: the MAAP4/VVER for reactor operation and shutdown mode with closed reactor vessel and the MELCOR code for shutdown mode with open reactor vessel. In this paper an overview of the Level 2 PSA methodology; description of the confinement; the interface between the level 1 and 2 PSA and accident progression analyses are presented. An evaluation of the confinement failure modes and construction of the confinement event trees as well as definition of release categories, source term analysis and sensitivity analyses are also discussed. The presented results indicate that: 1)for the full power operation - there is an 25% probability that the confinement will successfully maintain its integrity and prevent an uncontrolled fission product release; the most likely mode of release from the confinement is a confinement bypass after SGTM with conditional probability of 30%; the conditional probability for the confinement isolation failure probability without spray is 5%, for early confinement failure at the vessel failure is 4%, for other categories 1% or less; 2) for the shutdown operating modes - the shutdown risk is high for the open reactor vessel and open confinement; important severe accident sequences exists for release categories: RC5.1, RC5.2 and RC6.2
The gradual reconstruction of the Bohunice V1 nuclear power plant (Slovakia) represents the most extensive reconstruction of a nuclear power plant in operation as implemented worldwide up to now. Extensive reconstruction works in both civil construction and process parts, in instrumentation and control part, and in electric part enhanced both nuclear safety and operational reliability of Bohunice V1 in a significant way.(author)
During the refueling at the first unit of Bohunice NPP in 2005 a lot of sediment was found on the upper storage rack. This sediment was identification as a filter resin. Resin was found in most of the fuel assemblies, pipes and tanks of the primary circuit and his auxiliary systems. Resin producer and WANO network was contacted in order to get information about similar events. Management of Bohunice NPP made a decision that primary circuit, fuel assemblies and auxiliary systems have to be cleaned. Subsequent cleaning extended outage by 31 days. This paper summarizes causes, existing consequences and corrective actions. Accent was put on the hydraulic characteristics of the primary circuit measurement, power distribution core monitoring and the primary circuit water quality verification (Authors)
From October 2003 until december 2014 the Consortium consisting of Iberdrola Engineering and Construction (leader). Empresarios Agrupados Internacional, and Indra Sistemas has carried out the project Project Management Unit ((PMU) for the decommissioning of Bohunice V1 NPP (units 1 and 2), type VVER-440/V-230 in Slovakia. during the first phase (2003-2007) EdF was also part of the Consortium. The project is funded by the Bohunice International Decommissioning Support Fund (BIDSF) administered by the RBRD. The main objective of the project is to provide the necessary engineering and resources of project management for planning, execution, management, coordination and monitoring of all tasks in support of the decommissioning. (Author)
Experimental corrosion studies on non standard austenitic SS, A1, have been carried out. The samples were immersed in reactor coolant water medium with pH variation of 5.95, 6.0, 6.1, and 6.31. The experiments were carried out using a type of M-273 EG&G potentiostate/galvanometer test instrument. The post-corrosion samples' microstructure were analyzed with the aid of EDS (energy dispersive spectroscopy) equipped SEM instrument to detect the presence of any viable corrosion products. For further verification x-ray diffraction method was also used to detect any possible emerging corrosion products type on the samples' surfaces. Experimental results confirm that non standard austenitic SS immersed in reactor coolant water corrosion medium with a variation of concentration experience very little or almost no corrosion, and that according to the so-called Fontana's criteria these test-materials turn out to have an excellent resistance toward reactor coolant water corrosion medium. This is also evidenced by the very low corrosion rate value measured in this study. EDS study and X-ray diffraction results indicate that the possible ensuing corrosion by products are chrome oxides and iron oxides. (author)
This paper deals with project development of training programmes for non-licenced NPP personnel-masters, field operators, maintenance and technical supporting personnel. The programme development focuses on the part stage and on the job training at NPP. Bohunice NPP belongs to plants with higher specific number of personnel per installed power capacity. This factor also influenced the choice of programmes design. Undermentioned procedure is one of various approaches to SAT exploitation for training programmes design. (author)
The SCORPIO-VVER reactor core monitoring system is an advanced redundant software system without actuating members falling in the BT3 class which has been installed at the four Dukovany reactor units and at two units of the Slovak Jaslovske Bohunice V2 NPP. The system is described in detail and its history and experience gained at Dukovany are highlighted. (orig.)
The objective of the project was to accumulate technical experience with application of modern control theory in nuclear power by carrying out a case study of an experimental direct digital control at the A1 reactor about its nominal steady state. The research has proved that slightly modified method of solution of the linear stochastic regulator problem can be successfully applied in design of digital control system of nuclear power reactors
In this paper results of monitoring of chemical gaseous and liquid effluents into the rivers Vah River as well as of radiation monitoring of Bohunice V-2 NPP are presented. The radioactive effluents into atmosphere within August 2008 were: 2.72 MBq of aerosols, 0.052 MBq of of iodine-131, and 537 GBq of rare gases. For the period January - August 2008 these radioactive effluents into recipient of Vah River represented 6.66 MBq (0.318% of annually limit (AL)) of aero-soles; 0.318 MBq (0.00049% of AL) of iodine-131 and 3341 GBq (0.172% of AL) of rare gases. The radioactive effluents into Vah River recipient within August 2008 were: 968.78 GBq of tritium and 1.35 MBq of corrosive and fission products. For the period January - August 2008 these radioactive effluents into Vah River recipient represented 3,233.51 GBq (16.1675% of AL) of tritium, and 13.50 MBq (0.104% of AL) of corrosive and fission products. Average dose equivalents rate for the villages in surroundings of the Bohunice V-2 NPP for the period from July 31 to September, 2008 are published. Chemical effluents are also reported
The report contents final environment evaluation and selected characteristic of residents health physics of nuclear power plant Bohunice region. Evaluated data were elaborated during analytical period 1993-1997.Task solving which results are documented in this final report was going on between 1996- 1998. The report deals in individual stages with the following: Information obtaining and completing which characterize demographic situation of the area for the 1993-1997 period; Datum obtaining and completing which contain selected health physics characteristics of the area residents; Database structures for individual data archiving from monitoring and collection; Brief description of geographic information system for graphic presentation of evaluation results based on topographic base; Digital mapping structure description; Results and evaluation of radionuclide monitoring in environment performed by Environmental radiation measurements laboratory by the nuclear power plant Bohunice for the 1993-1997 period. Demographic situation evaluation and selected health physics characteristics of the area of nuclear power plant residents for the 1993-1997 period are summarized in the final part of the document. Monitoring results and their evaluation is processed in graph, table, text description and map output forms. Map outputs are processed in the geographic information system Arc View GIS 3.0a environment
Zhang, Hong; Xu, Xuebing; Mu, Huiling; Nilsson, J.; Adler-Nissen, Jens; Høy, Carl-Erik
Lipozyme IM-catalyzed interesterification of the oil blend between palm stearin and coconut oil (75/25 w/w) was studied for the production of margarine fats in a 1 kg scale batch stirred tank reactor. Parameters such as lipase load, water content, temperature, and reaction time were investigated...
Zhang, Hong; Xu, Xuebing; Mu, Huiling;
Lipozyme IM-catalyzed interesterification of the oil blend between palm stearin and coconut oil (75/25 w/w) was studied for the production of margarine fats in a 1 kg scale batch stirred tank reactor. Parameters such as lipase load, water content, temperature, and reaction time were investigated...
Annual impacts of Bohunice V -2 operation caused by effluents in routine releases during the year 2007 were calculated and evaluated first time with the help of completely new program - ESTE AI. Program is approved by the 'Public Health Authority of the Slovak Republic' and since January 2008 is used as legal instrument by Slovenske elektrarne a.s., NPP Bohunice. In this poster presented are annual effluents to the atmosphere and to the hydrosphere from Bohunice V-2 NPP. Presented, analyzed and discussed are main results of 2007 impacts evaluation. (authors)
The main goals of actual stage of this long-term project are: • Analyses of the seventh set of specimens No.116/2 from ATABOR steel after 7, 5 eventually 4 years of exposition in the Bohunice spent fuel interim wet storage environment using the following methods: o Documentation of sample surfaces after removal from storage pool; - Microstructure evaluation for the base and weld metal; - Analysis of corrosion media influence to the structure of ATABOR steel using the light microscopy; o Scanning electron microscopy and microanalysis of exposed samples; • The optimization of database structure on the base of experience from the second stage of SPAR-III contract; • On the basis of the analysis results to prepare the new optimized set of specimens prepared from the specific heat of ATABOR steel
Richter, D; Tostevin, G; Skrdla, P
Results of thermoluminescence (TL) dating of 11 heated flint artifacts from the 2002 excavation at Brno-Bohunice, Czech Republic, are presented. The samples are from the eponym locality for the Bohunician, an industrial type considered technologically transitional between Middle and Upper Paleolithic core reduction strategies. The Bohunician is the first early Upper Paleolithic technocomplex in the Middle Danube of Central Europe and, therefore, is implicated in several issues related to the origins of modern humans in Europe. The Bohunician provides an example of how one technological strategy combines crested blade initiation of a core with the surficial (almost Levalloisian) reduction of blanks as blades and points. As the Middle Danube lacks antecedents of the behavioral steps within this technology, several hypotheses of inter-regional cultural transmission, with and without hominin gene flow, could explain the appearance of the Bohunician. The elucidation of the temporal context of Bohunician assemblages is, therefore, a critical step in understanding the behavioral, and potentially biological, succession in this region. Radiocarbon age estimates from charcoal associated with Bohunician sites suggest a wide age range between 33 and 41 ka 14C BP, which is also observed for individual sites. TL dating of heated flint artifacts provides ages on the calendric time scale of an archeological event, the firing. The weighted mean of 48.2+/-1.9 ka BPTL for 11 heated flint samples from Brno-Bohunice provides the first non-radiocarbon data on archeological material from the Bohunician. The TL dating, in conjunction with the archeological and sedimentological analysis, allows the evaluation of the integrity of this new type-collection. The hypothetical possibility of the incorporation of Szeletian artifacts (i.e., leaf points) into the site formation processes can therefore be refuted. PMID:18951613
Over the past 20 yr, the number of nuclear reactors on university campuses in the United States has decreased from >70 to <40. Contrary to this trend, the University of Texas at Austin recently completed construction of a new reactor facility at a cost of $5.8 million. The TRIGA Mark II reactor in this facility will be licensed for 1.1-MW steady-state operation and $3.00 power-pulse transients. The new reactor facility was established to enhance the instructional and research opportunities in nuclear science and engineering for both undergraduate and graduate students at the University of Texas. In addition to neutron activation analysis, programs are being planned and equipment is being designed for neutron depth profiling, prompt gamma activation analysis, neutron radiography, and cold neutron research. Because of continued interest in fission-yield system developed by the author when he was at the University of Illinois. The operation of this unique system for the direct physical measurement of independent yields in thermal-neutron fission is reviewed in this paper
Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data
In this leaflet results of exploitation of four units of the Bohunice V-1 and V-2 NPPs are presented. The electricity and heat production in February 2005 are reviewed. Within a February 2005 the electricity was produced: 217 GWh (block 1), 281 GWh (block 2), 277 GWh (block 3), 282 GWh (block 4), totally 1057 GWh, and 2271 GWh within a January - February 2005. The heat production in February 2005 was 266 506 GJ, and within a January - February 2005 it was produced 531 849 GJ of heat. On February 17 Slovak minister of economy Pavol Rusko and general director of ENEL Paolo Scaroni signed the agreement on acquisition of 66 per cent of Slovenske elektrarne (SE) by Italian ENEL for 840 million Eur. SE has capacity of around 7 GW (83 per cent of total Slovakian capacity). In 2004 SE generated 26 TWh of electricity. Processing and storage of radioactive wastes in Decommissioning of Nuclear Installations and Spent Fuel and Rad-waste Management (SE-VYZ) is presented. Since beginning of this year 58 fibre-concrete containers have been filled up in Bohunice processing centre of radioactive wastes. Twenty-three pieces of fibre-concrete containers were processed into fibre-concrete containers in Bohunice processing centre of radioactive wastes (BSC RAO) in February 2005. Twenty fibre-concrete containers were stored into Republic storage of radioactive wastes (RU RAO). Total number in RU RAO reached 830 pieces of fibre-concrete containers, which represent 11.53 per cent of storage capacity (7200 containers). Bohunice processing centre of radioactive wastes was put into active operation just before five years
We pursue the vicinity of Bohunice NPP. The vicinity has cyclic form with radius of 30 km, what represents an area approximately 2 800 km2. This area of pursued vicinity is requisite by the security report of Bohunice NPP. To the presumptive calculations we used the complete databases of Register of death, Register of municipalities and of Register of age structure of the inhabitants of the Slovak republic from 1993 to 1999, fully-fashioned in Statistical authority of the Slovak republic. We work with databases, which don't contain personal identifications. We pursue the evolution of the mortality by the indicators of the mortality, calculated by the WHO. By the literary sources and by our experience is necessary the sum at least of three years to calculation of stable demographic and epidemiological parameters. Therefore we work with the method of short time series. The basic observed unit, which is represented by one value of the indicator, is one municipality. All our assessing analyses are calculated from triennial sums of all indicators, so we work with man-years. Advanced report is the adjusted extract from Complex report on situation of environment and health of the inhabitants in vicinity of Bohunice NPP in 1999, which was advanced by our society in March 2001. (authors)
Endo, S; Yoshitake, Y
The energy spectrum of fission neutrons in the biological irradiation field of the Kinki University reactor, UTR-KINKI, has been determined by a multi-foil activation analysis coupled with artificial neural network techniques and a Au-foil activation method. The mean neutron energy was estimated to be 1.26+-0.05 MeV from the experimentally determined spectrum. Based on this energy value and other information, the neutron dose rate was estimated to be 19.7+-1.4 cGy/hr. Since this dose rate agrees with that measured by a pair of ionizing chambers (21.4 cGy/hr), we conclude that the mean neutron energy could be estimated with reasonable accuracy in the irradiation field of UTR-KINKI. (author)
For the integrity evaluation of steam dryers in up-rated nuclear power plants, we have applied acoustic analysis to a nuclear power plant steam dome and main steam pipings. We have selected a 1,100 MWe-class boiling water reactor as a subject of the analysis. We have constructed a three-dimensional finite element model, and conducted acoustic analyses. The analysis result suggested that the origin of steam pressure pulsation in high frequency range was due to vortex shedding at standpipes. (authors)
Zhang, Hong; Xu, Xuebing; Mu, Huiling; Nilsson, Jörgen; Adler-Nissen, Jens; Høy, Carl-Erik
Lipozyme IM-catalyzed interesterification of the oil blend between palm stearin and coconut oil (75/25 w/w) was studied for the production of margarine fats in a 1 kg scale batch stirred tank reactor. Parameters such as lipase load, water content, temperature, and reaction time were investigated...... higher temperature. Addition of water to the enzyme increased the contents of diacylglycerols and FFA in the products linearly. However, it had no effect on the degree of interesterification for the first batch when the enzyme was reused. Lipozyme IM was stable in the 10-batch test after adjusting the...
In this paper results of monitoring of chemical gaseous and liquid effluents into the Vah River as well as of radiation monitoring of Bohunice V2 NPP are presented. The radioactive effluents into atmosphere within January 2008 were: 0.19 MBq of aerosoles, 0.025 MBq of iodine-131, and 382 GBq of rare gases. For the period January 2008 these radioactive effluents into atmosphere represented 0.190 MBq (0.0002% of annually limit (AL)) of aerosoles, 0.025 MBq (0.00004 MBq of AL) of iodine-131, and 382 GBq (0.519% of AL) of rare gases. The radioactive effluents into recipient of Vah River within January 2008 were: 0.06 GBq of tritium, and 0.63 MBq of corrosive and fission products. For the period January 2008 these radioactive effluents into recipient of Vah River represented 0.06 GBq (0.0003% of AL) of tritium; 0.63 MBq (0.005% of AL) of corrosive and fission products. Dose equivalents rate for the villages in surroundings of the Mochovce NPP in January 2008 are published. Chemical effluents are also reported.
In this paper results of monitoring of chemical gaseous and liquid effluents into the Vah River as well as of radiation monitoring of Bohunice V2 NPP are presented. The radioactive effluents into atmosphere within May 2008 were: 0.29 MBq of aerosoles, 0.042 MBq of iodine-131, and 367 GBq of rare gases. For the period January - May 2008 these radioactive effluents into atmosphere represented 1.10 MBq (0.0014% of annually limit (AL)) of aerosoles, 0.159 MBq (0.00024% MBq of AL) of iodine-131, and 1.930 TBq (0.519% of AL) of rare gases. The radioactive effluents into recipient of Vah River within May 2008 were: 150.68 GBq of tritium, and 0.38 MBq of corrosive and fission products. For the period January - May 2008 these radioactive effluents into recipient of Vah River represented 855.43 GBq (4.2772% of AL) of tritium; 2.92 MBq (0.0226% of AL) of corrosive and fission products. Dose equivalents rate for the villages in surroundings of the Mochovce NPP in May 2008 are published. Chemical effluents are also reported.
In this paper results of monitoring of chemical gaseous and liquid effluents into the Vah River as well as of radiation monitoring of Bohunice V2 NPP are presented. The radioactive effluents into atmosphere within February 2008 were: 0.19 MBq of aerosoles, 0.025 MBq of iodine-131, and 344 GBq of rare gases. For the period January - February 2008 these radioactive effluents into atmosphere represented 0.37 MBq (0.0005% of annually limit (AL)) of aerosoles, 0.050 MBq (0.00008% MBq of AL) of iodine-131, and 725 GBq (0.036% of AL) of rare gases. The radioactive effluents into recipient of Vah River within February 2008 were: 100.01 GBq of tritium, and 0.61 MBq of corrosive and fission products. For the period January - February 2008 these radioactive effluents into recipient of Vah River represented 110.07 GBq (0.5504% of AL) of tritium; 1.24 MBq (0.010% of AL) of corrosive and fission products. Dose equivalents rate for the villages in surroundings of the Mochovce NPP in February 2008 are published. Chemical effluents are also reported.
Zhang, Hong; Xu, Xuebing; Mu, Huiling;
Lipozyme IM-catalyzed interesterification of the oil blend between palm stearin and coconut oil (75/25 w/w) was studied for the production of margarine fats in a 1 kg scale batch stirred tank reactor. Parameters such as lipase load, water content, temperature, and reaction time were investigated...... interesterification. A Lipozyme IM load of 6% was required for a reaction of 6 h and at 60 °C, to reach a stable degree of interesterification. Temperature variation in the range of 50–75 °C did not affect the reaction degree as well as the contents of diacylglycerols, but the content of FFA slightly increased with...... higher temperature. Addition of water to the enzyme increased the contents of diacylglycerols and FFA in the products linearly. However, it had no effect on the degree of interesterification for the first batch when the enzyme was reused. Lipozyme IM was stable in the 10-batch test after adjusting the...
Bohunice V1 in Slovakia is a two unit WWER 440/230 whose units went into commercial operation in 1979 and 1981 respectively. The plant was not initially designed for seismic loading. Later geotechnical studies concluded that the site seismic hazard should be defined as an earthquake of MSK 8 intensity. This relates to approximately 0.25 g peak ground acceleration in the free field at the site. Some early reconstruction to strengthen the plant against earthquakes was done in the early 1990s but did not include all safety significant structures and equipment. In 1996, EBO, the plant operator, entered into a contract with consortium REKON, a Siemens and VUJE joint venture, for a major reconstruction program to update all safety systems required for a safe shutdown, to improve integrity of confinement and assure spent fuel cooling. This reconstruction project includes verification of seismic adequacy of all safety related structures and equipment in the REKON scope which is not being replaced by new construction. Siemens and EQE International are jointly conducting the seismic verification and required upgrading for the existing structures and equipment. Criteria for the verification and upgrading were developed for the project utilizing Technical Guidelines provided by IAEA, Reference 1, and linking them with international and local codes and standards and specific methodologies developed for similar projects in the US and Western Europe. The criteria are briefly discussed herein and are summarized in a companion paper, Reference 4. Because of the major improvements being implemented in safety systems, much of the essential safety related equipment is being directly replaced or completely new systems are being constructed that supersede existing ones. Consequently, a significant amount of the equipment that would normally require seismic adequacy verification is deleted from the verification scope (see Table 4). The reconstruction project will continue through 1999
A series of experiments is under way at the Karlsruhe fast zero power reactor SNEAK for the investigation of steam-cooled fast reactors in the 100-MW(e) range. This series started in May with the critical experiment of SNEAK 3A-1, a 670-litre uranium system containing 7.41 x 1020 atoms/cm3 of hydrogen in the form of polyethylene foils. The neutron physics of this assembly has been studied in detail. The neutron energy spectrum has been measured by various methods from the eV-region to more than 1 MeV in the core centre and at the periphery, reaction rates have been measured in the centre and in axial and radial traverses, and the initial breeding ratio and the reactivity worth of selected materials have been determined. Measurements of the Doppler reactivity effect, the steam void effect and of β/ℓ have been performed. Special attention has been given to the experimental investigation of heterogeneity effects. The experimental results are compared with calculations using the 26-group ABN set and a specially prepared 26-group cross-section set KFK-SNEAK using latest cross-section information and the SNEAK-3A spectrum as a weighting spectrum. The heterogeneity results are compared with theoretical models including space-dependent resonance self-shielding. The series of SNEAK-3 experiments is now being continued with the uranium Assembly 3A-2, which has about twice the hydrogen concentration of 3A-1. After the measurements in this system have been completed the inner part of the core will be replaced by an equivalent plutonium-fuelled zone thus forming the two-zone core SNEAK-3B. (author)
Ješkovský, Miroslav; Povinec, Pavel P.; Steier, Peter; Šivo, Alexander; Richtáriková, Marta; Golser, Robin
Atmospheric radiocarbon has been monitored around the Nuclear Power Plant (NPP) Jaslovské Bohunice (Slovakia) using CO2 absorption in NaOH solution since 1969. In 2012, tree ring samples were collected from Tilia cordata using an increment borer at Žlkovce monitoring station situated close to the Bohunice NPP. Each tree ring was identified and graphite targets were produced for 14C analysis by accelerator mass spectrometry. The 14C concentrations obtained from the tree-ring samples have been in a reasonable agreement with the averaged annual 14C concentrations in atmospheric CO2.
In this leaflet the results of monitoring of chemical gaseous and liquid effluents into the Vah River and Dudvah River as well as of radiation monitoring of Bohunice V1 NPP, Interim Spent Fuel Storage (MSVP), Bohunice Radioactive Waste Processing Centre (VK808 - BSC), The Main Production Unit (VK 46A - HVB) and Bitumenation Lines (VK 46B - BL) are presented. The radioactive effluents into atmosphere within January - May 2013 (for NPP V1, MSVP, BSC, HVB and BL, respectively) were: 0.138 MBq (0.000% of AL) for V1, 0.194 MBq (0.065 of AL) for MSVP, 0.192 MBq (0.136% of AL) for BSC, 0.470 MBq (0.071% of AL) for HVB and 0.168 MBq (0.119% of AL) for BL of aero-soles. The radioactive effluents into atmosphere and hydrosphere within a May 2013 for NPP V1, MSVP, and VK 808 BSC, respectively, were: 0.024 MBq (V1), 0.055 MBq (VK MSVP), 0.070 MBq (VK 808 BSC), 0.074 MBq (VK 46A) and 0.011 MBq (VK 46B) of aero-soles into atmosphere; 1.249 MBq (V1 and MSVP) and 6.744 MBq (TSU RAO and NPP A1) of corrosive and fission products, and 5.723 GBq (V1 and MSVP) and 5.744 GBq (TSU RAO and NPP A1) of tritium into the Vah River and 0.000 GBq (V1 and MSVP) and 0.000 MBq (TSU RAO and NPP A1) of corrosive and fission products and 0.000 GBq of tritium (V1) and 0.000 GBq of tritium (TSU RAO and NPP A1) into the Dudvah River). For the period January - May 2013 these radioactive effluents into recipient of Vah River represent for corrosive and fission products 5.653 MBq (0.043% of AL for V1 and MSVP) and 34.953 MBq (0.291% of AL for TSU RAO and MSVP); and for tritium it is 8.397 GBq (0.420% of AL) for V1 and MSVP, and 31.661 GBq (0.317% of AL) for TSU RAO and NPP A1 (into the Vah River) and for corrosive and fission products 0.000 MBq (V1 and MSVP) and 0.000 MBq (TSU RAO and NPP A1) (0.000% of AL) and 0.000 GBq (0.000% of AL) for V1 and MSVP and 0.000 GBq (0.000% of AL) (TSU RAO and NPP A1) of tritium (into the Dudvah River). Chemical effluents are reported, too.
In the presentation, some information concerning the historical background of NPP A-1 in Jaslovske Bohunice, Slovakia is given. The main technical parameters used during production activities concerning the decommissioning of the NPP A-1 to a first stage (i.e. to obtain radiologically safe stage) are solved together with the main contractor, Nuclear Power Plant Research Institute, Trnava, according to an approved project by the Slovak Government and Nuclear Authorities. The technological schemes for the radioactive waste treatment at SE-VYZ o.z. and their main technical parameters are shown as well. (author)
Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)
The degradation of reactor pressure vessel (RPV) steel is a complex process depending on many factors (thermal and radiation treatment, chemical compositions, preparing conditions, ageing, operation environment, etc.). This paper describes tests based on Nondestructive Methods used for evaluation of material characterisation at Slovenske Elektrarne a.s. Positron Annihilation Spectroscopy (PAS), Moessbauer Spectroscopy (MS) and Transmission Electron Microscopy (TEM) investigate microstructure changes of Reactor Pressure Vessel steels caused neutrons irradiation. There are showed results of investigation of reactor pressure vessel steel specimens after five years irradiation in reactor nature by mentioned NDT methods. Investigated specimens has been prepared for the Extended surveillance specimen program which has run out on the 3rd and 4th units of NPP Jaslovske Bohunice and for the Modified surveillance specimen program in the 1st and 2nd unit which is continued in NPP Mochovce (Slovakia). PAS and MS spectra showed that the degradation of the steel properties associated with the effects of neutron irradiation can be well detected. The samples from RPV base metal (15Kh2MFA) and weld metal (Sv 10KhMFT) were measured by PAS and MS before and after irradiation. Samples have been irradiated in VVER-440 reactor (units 3rd and 4th in Bohunice as well as 1st and 2nd units Mochovce) by neutron fluency from 7.8 1023 m-2 up to 2.5 1024 m-2. Measurement results are presented and discussed in detail.(author)
The first pilot NPP in the former Czechoslovakia was A1, which was built at Jaslovske Bohunice near the town of Trnava. An NPP with a capacity of 143 MW(e), it was commissioned in 1972 and operated with interruptions until 1977. A KS-150 reactor with natural uranium as fuel, D2O as moderator and gaseous CO2 as coolant was installed in the plant. The first serious accident associated with refuelling occurred in 1976, when a locking mechanism at a fuel assembly failed. The core was not damaged during that accident and after reconstruction of the damaged technology channel, the plant resumed operation. The second serious accident (level 4 according to the International Nuclear Event Scale) occurred in 1977, when a fuel assembly overheated, causing release of D2O into the gas cooling circuit. This accident was attributed to human error during replacement of a fuel assembly. Subsequent rapid humidity increase in the primary system resulted in damage to fuel elements in the core, and the primary system was contaminated by fission products. Internal reactor structures were also damaged. Radioactive contaminants penetrated into parts of the secondary system by leaking through steam generators. The radiation impact on and around the plant site was below specified limits for both events. Based on a technical and economic study of the difficult equipment repairs needed to restore plant operation, and also due to the policy decision to discontinue further construction of gas cooled reactors in the former Czechoslovakia, a decision was made in 1977 to terminate plant operation. The decision to proceed with the A1 plant decommissioning was issued in 1979. Beginning in 1981, decommissioning proceeded with disassembly of equipment from the secondary system (process equipment in the machine hall, turbines with auxiliaries, feed water tanks, diesel generator station, pumps, cooling towers, electric equipment). At the same time, other systems were disassembled, which included turbine
Vertical displacements were measured of the foundations and of selected bearing structures of the V-1 nuclear power plant buildings during the plant's construction and operation. Measured were displacements of the engine room foundations, the reactor building, the boron management building, the turbogenerator building, the cooling towers, the ventilation stack, and the foundations of buildings showing adverse properties. Some results are presented. (E.J.). 4 figs., 2 refs
The implementation of the automated control system at the Jaslovske Bohunice nuclear power plant is taking place in two stages. In the first stage the main computer centre was built with an ES 1055 M computer which has been in operation since January 1985. In the following year, 6 local network terminals of type ES 7927 were installed and one multiplexer ES 8371 for the control of the network. All the equipment operates reliably, the ES 7039 printers are not so reliable. The weakest element are large-size magnetic disc memories. In the next stage, the construction is envisaged of a terminal network with SM 4-20 and SM 52/11 computers and 7202 terminals. The contribution of the implementation of computer technology so far has been in the field of maintenance where it has allowed to centrally plan repairs of some 15,000 items and to coordinate the activities of sub-contractors. Also positive are results in the field of measurement and control technology where the reliability is being evaluated of some 20,000 measurement circuits and elements, and their preventive maintenance and repairs planned. Briefly summed up are items for further increasing the contribution of the deployment of computer technology in nuclear power plants. (Z.M.)
The study was apart of investigation of the size distribution of aerosol in air effluents from NPP V1 Jaslovske Bohunice. The evaluation the possible relationship between aerodynamic diameter of aerosol particles and chemical forms of radionuclides attached to the discharged aerosol was tried. Selective leaching was used for speciation of radionuclides present in the aerosol particles and for the estimation of their behaviour in the environment and absorption in gastro-intestinal tract. Activity concentrations of the radionuclides in the air, collected on collection substrates taken from individual impact stages and on back-up filter, were determined by sensitive gamma-spectrometric analysis using high purity Ge detectors. For the individual groups seven leaching steps were used. Following 12 radionuclides: silver-110m, cobalt-58, cobalt-60, cesium-134, cesium-137, manganese-54, ruthenium-103, antimony-124, antimony-125, tin-113, zinc-65, zirconium-95. Result shows that the leached fraction of the of the activity concentration does not depend on the size of the aerosol particles. (J.K.) 3 tabs., 3 figs
Results are presented of welding and detailed evaluation of test plates and rings of actual sizes, and the results are related directly to the determination of the thermal conditions of the welding of the A1 steel reactor vessel. Electro slag welding and automatic C02 welding were used. Details are given of the materials used and the conditions of the tests. Test results and analysis are given under the following heads: thermal cycles, microstructure, notch toughness; strain effect, stress state, specific accumulated energy; temperature of brittle crack initiation, the critical size of the initial flaw; conclusions. (U.K.)
This work shows the approach used to the numerical simulation of the thermal test of a 1:2 scale model of a dual purpose cask (transportation and/or storage) for spent fuel elements from nuclear research reactors. Conservatively, the cask impact limiters are not modeled. This test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. Also, it is part of an IAEA sponsored project which includes Latin American countries with research reactors. This cask model has a stainless steel double wall cylinder (which contains the biological lead shielding) with flat heads and internal structures to accommodate the fuel elements. The cask project is described briefly as well as the developed finite element model and the main adopted hypothesis to consider the non-linearities as thermal contacts, properties varying with the temperature, phase change (thermal shielding lead) using the enthalpy method, and radiation among the internal parts. The analysis will cover the 30 min heating condition at 800 deg C and about 2 hours of the cooling phase. As the main purpose of the paper is to present the proposed approach for the thermal test numerical simulation, only some preliminary numerical results are shown without any comparison to the experimental ones. (author)
In this chapter two serious reactor accidents in the A1 nuclear power plant in Jaslovske Bohunice (Slovak Republic) are described. The first accident - during replacement of the fuel assembly the fuel element went off from the reactor on January 1976. Residual power of reactor at the time of the fuel assembly outspreaded from the channel H05 (January 5, 1976, 11:55 hours) was determined at 0.63% of the nominal power of the reactor where he worked before shutdown, i. e. residual power of the reactor was about 2.9 MWt. Leakage of carbon dioxide was stopped by loading machine. Maximum temperature 565 grad C was registered by measurement of uranium temperature. As it appeared later, overheating damaged the cover of several fuel assemblies in the central zone and the inner peripheral border zone. During the accident power radiation situation grew worse in most areas of plant. Released radioactive gases caused the maximum effective dose 10-8 Sv per capita in the territory to a distance of 25 km from the plant. At that time permitted individual dose per individual of the population was 5 mSv/year. The consequences of the accident on technological equipment are analyzed. Dealing with the aftermath of the accident on the technological equipment is described. Planned replacement of spent nuclear fuel in a reactor in the technology channel C05 started on 22 February 1977 in the afternoon. The exchange was taking place in the reactor operation. The electrical power was 93 MW during refuelling. Exchange by filling machine took place normally. Launching of the fuel assembly from cool zone into the reactor core came after. In this operation, which began at about 18:13 pm fuel assembly was overheated. Consequently, the influence of high temperature induced its destruction, which caused damage of 'decompression' heavy water tube container. Big leaking was formed and heavy water began to penetrate into the gas of the primary circuit. The operator immediately stopped the reactor
Positron annihilation lifetime measurements using the pulsed low energy positron system (PLEPS) were applied for the first time for the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in the microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation thermal treatment can be detected. The samples originated from the Russian 15Kh2MFA and Sv10KhMFT steels, commercially used at WWER-440 reactors, were irradiated near the core at NPP Bohunice (Slovakia) to neutron fluences in the range from 7.8x1023 to 2.5x1024 m-2
Sipping equipment for the Soviet-type VVER-440 pressurized water reactors was developed on the basis of the proven in-core sipping technique used for boiling water reactor fuel assemblies. The main components of the system are the sipping hood with seven test positions, the control panel for system operation and sample collection, and the manifold connection line. During testing the upper ends of the hexagonal fuel assemblies are lifted into the air-filled sipping hood to interrupt the coolant flow by means of pneumatically actuated grippers. The first equipment of this kind has been in use in the nuclear plant Jaslovske-Bohunice, Czechoslovakia, since 1986. (orig.)
The ''fault tree'' method and its application in the reliability analysis of the control and safety system of the A-1 nuclear power plant is described. The fault tree is a logical model involving all probable fault combinations of components and subsystems associated with the occurence of the final undesired event - the system failure. The method makes possible a quantitative and qualitative analysis of the system reliability and availability using a digital computer. The aim of the fault tree reliability analysis is to determine the distribution of reliability in the system, to find ''weak spots'' of the system considered and to define minimum repair times for the system components. (author)
Mineralogical characterization of a soil from the Nuclear Power Plant Jaslovske Bohunice object 41 and 44/10 was performed by x-ray diffraction method and surface measuring. It was find that the main component was smectite but other minerals like quartz, illite, chlorite and K-feldspar, were observed. Sorption and speciation of 137Cs, 85Sr, 241Am and 60Co in the mix soil was done with continuous leaching and Tessier scheme. It was found that distribution ration for Co and Cs is the highest. Strontium was in ionic form, Cs was leached with nitric acid, americium was fixed on carbonate fraction and cobalt was predominantly leached with reduction solutions as can be observed from the Tessier sequence method. Sorption of iodine on the soil used in the experiment was negligible (authors)
Project of decommissioning NPP A-1 is split into 4 main groups of tasks. Tasks in group 1 are focused on the solution of selected problems that have immediate impact on the environment. It is mainly the solution of problems in the building of cleaning station of wastage water and in the building with underground storage tanks for wastage water and solid radwaste, including the prevention of wash-out and penetration of contaminated soil from these buildings into surface and underground waters. A part of addressing these tasks is a controlled of generated radwaste-predominatly sludge with various physical and chemical properties. Tasks in group 2- following the removal of spent fuel-are focused on the management of all radwaste in the long-term storage facility, in the short-term storage facility, equipment of transport and technology part, equipment in hot cells. Tasks in group 3 are focused on development of technology procedures for treatment and conditioning of sludge, contaminated soils and concrete crush, saturated ionexes and ash from incineration facility of the Bohunice radwaste treatment and conditioning complex. Tasks in group 4 are focused on the methodology. And technical support for particular activities applicable during decommissioning NPP
The nuclear power plant V-1 Bohunice consists of two VVER-440 units with V-230 reactors. Unit 1 was commissioned in 1978 and Unit 2 in 1980. Large experience and knowledge from the operation of previous units with V-230 reactors were incorporated into the V-1 design, which resulted in a higher level of safety and operational reliability of these units. The Siemens company which won an international bidding process developed these basic goals for the Gradual Upgrading into the so called Basic Engineering (BE). For the implementation of the Gradual Upgrading in line with the BE, Rekon consortium was established consisting of Siemens and VUJE. The implementation of the Gradual Upgrading is scheduled for the time period of 1996 - 2000. Siemens was responsible for the upgrading strategy - based on the approved results of the basic engineering phase and the PSAR, the engineering and realization of all I and C improvements, and also for the seismic upgrade. VUJE's responsibility covered the detailed engineering and implementation of mechanical, electrical and civil part of upgrading measures as well as overall organisation and evaluation of verification tests. The consortium awarded contracts for final planning and design, installation services and commissioning to other Slovakian subcontractors in order to ensure the largest possible local content. The gradual reconstruction of the V-1 Bohunice with V230 reactors represents a comprehensive reconstruction of safety-related systems and equipment. Following its completion, the units will be operated with a safety level accepted internationally. (author)
An overview is given of the recommendations and regulations concerning IandC systems of nuclear power plants with WWER-440 reactors in the Slovak Republic, and of the relevant involvement of regulatory bodies. The issues included in regulatory decrees and pertaining to the safety aspects of IandC systems are mentioned point by point, with emphasis put on the upgrading of the systems. The power plants covered include the Bohunice V-1 and V-2 plants and the Mochovce-1 and Mochovce-2 units. (A.K.)
Regulatory authority decided in 1993 that will use REG 1.70 format for FSARs of Bohunice NPP. The new format first is used for V-2 FSAR in 1995 (upgrading after 10 years of commercial operation). After issuing of V-2 FSAR, it passed several internal review and IAEA mission. During the process of estimation of FSAR many shortcomings and problems is discovered. The main discovered general problems concerns sharing of architect-engineer role between domestic and foreign organisation, absence or unavailability of a part of the design basis information and missing of national guidelines and need to adopt international (IAEA) or foreign ones (US, German, French) guidelines for specific types of analyses). Difficulties of FSAR development are generated by application of different approaches to the design, construction (original design standards - OPB) and to the assessment of the safety of the NPP (western, IAEA), as well as a lack of communication between the NPPs / national engineering organizations and the General designer. The shortcomings of Chapter 15 is expressed in justification of the computational variants, consistency (data, modeling approach, assumptions), different approach to conservatism in individual subchapters (application of SFC, boundary conditions, initial conditions), acceptance criteria not always properly evaluated and 'author / user effect'. The conclusion of estimation of present situation is based on the implications and shows necessity for overall correction and extension of the FSAR (requirements surpassing REG 1.70). About Chapter 15 is achieved unification of the approach in different areas of AA (consistency) and elimination of inconsistencies, including of bounding scenarios into all subchapters and general format related improvements (legibility, quality of graphics etc.) Development of plant specific accident analysis methodologies are started in 1995 in collaboration with VUJE Trnava and main basis are the drafts of IAEA-WWER-EBP-01
The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping
The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.
This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.
Up to the TMI accident the water level inside the pressurizer was used to monitor the water inventory inside the primary cooling system of pressurized water reactors. The TMI accident showed that this was not a reliable measurement for the reactor coolant inventory inside the reactor pressure vessel. For this reason there was a demand for a measurement of the water level inside the RVP, independent from the existing one inside the pressurizer and with a diverse measuring method. For WWER reactors a new level measurement system was developed to monitor the water level inside the reactor pressure vessel by means of the KNITU, resp. KITU level probe which meet all the mentioned engineered safeguards and geometric and constructive requirements. First backfitting s of the new level measurement system in the WWER s 440 in Bohunice V1 (Slovakia), unit 1 (1998) and unit 2 (2000), Novovoronezh (Russia), unit 4 (1999) and Kola (Russia), unit 1 and unit 2 (1999) show very good operational results. (Authors)
The prime objective of the IAEA Technical Co-operation Project on Evaluation of Safety Aspects of WWER-440 model 213 NPPs is to co-ordinate and to integrate assistance to national organizations in studying selected aspects of safety for the same type of reactors. Consequently, the study integrated the results generated by national activities carried out in the Czech Republic, Hungary, Slovakia and Ukraine and co-ordinated through the IAEA. Valuable assistance in carrying out the tasks was also provided by Bulgaria and Poland. A set of publications is being prepared to present the results of the project. The publications are intended to facilitate the review and utilization of the results of the project. They are also providing assistance in further refinement and/or extension of plant specific safety evaluation of model 213 NPPs. This Technical Document addressing the design basis and safety related design features of WWER-440 model 213 plants is the first of the series to be published. It is hoped that this document will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, 36 figs, tabs
Highlights: • Tetraselmis suecica production in a 1-ha GWP plant in Tuscany (Italy) has a NER < 1. • Major energy costs are embodied energy of GWP and mixing. • In a suitable location (North Africa) the NER increases by 40%. • Integration of photovoltaic in the GWP allows to achieve a NER of 1.7. • T. suecica cultivated in a GWP plant can yield up to 30 t of protein ha−1 year−1. - Abstract: The annual productivity of Tetraselmis suecica in a 1-ha Green Wall Panel-II (GWP-II) plant in Tuscany (Italy) is 36 t (dry weight) ha−1 year−1, which corresponds to an energy output of 799 GJ ha−1 year−1. The energy inputs necessary to attain that productivity amount to 1362 GJ ha−1 year−1, mainly given by the embodied energy of the reactor (about 30%), mixing (about 40%), fertilizers (11%) and harvesting (10%). The Net Energy Ratio (NER) of T. suecica production is thus 0.6. In a more suitable location (North Africa) productivity nearly doubles, reaching 66 t ha−1 year−1, but the NER increases only by 40% and the gain (difference between output and inputs) remains negative. In a GWP-II integrated with photovoltaics (PV), the NER becomes 1.7 and the gain surpasses 600 GJ ha−1 year−1. Marine microalgae cultivation in a GWP plant, in a suitable location, can attain high biomass productivities and protein yields 30 times higher than those achievable with traditional crops (soya). When the GWP reactor is integrated with PV, the process attains a positive energy balance, which substantially enhances its sustainability
Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...
This report describes all nuclear installations in the Slovak Republic. It informs the public about the safety of nuclear installations. The spent fuel activities and nuclear wastes storage matters are discussed separately ((NPP Bohunice V-1, NPP Bohunice V-2, NPP Mochovce, NPP Bohunice A-1, Radioactive wastes repository Mochovce, Interim spent fuel storage Bohunice)
A new concept and measurement procedures for clearance of building has been developed and implemented at NPP A1 (Bohunice). They are based on measurements of total RN activity per unity area on monitored surfaces at the erected building according to EU Recommendation RP-113 (limit of pure Cs-137 - 10 Bq/cm2). HD-RNs are taken into account in a standard way by summation formula supposing the known RN vector. The free release measurement procedures are complex, they include pre- and post- decontamination and free release measurements and the graded approach concerning measurement costs has been taken into account, as well. So the higher contamination potential of concrete structures is the more measurements has been done (more dense inhomogeneity mapping, sampling for scaling factors, etc.). Advantage of the new concept is that it avoids costly whole volume measurement of building rubble (e.g. 200 L drum monitor). It has been replaced by whole surface monitoring of the walls by portable LaBr spectrometer in 1 x 1 m grid. Detectors for this purpose must be metrologically certified by Slovak Metrology Institute. The new procedure has been firstly implemented at pilot clearance of four reservoirs (6 m in diameter, 4 m depth) in garden of object No. 41 in 2010-11. It is so late because it was not developed an effective free release procedure with reliable metrologically accepted clearance measurements. (author)
The Reactor Physics and MYRRHA 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 of reactor fuel. This expertise is applied in the 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 by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised
SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the 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 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised
Ait Abderrahim, A
The Reactor Physics and MYRRHA 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 of reactor fuel. This expertise is applied in the 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 by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.
After the earthquake and subsequent tsunami in March 2011 Fukushima nuclear power reactors in Japan were damaged. As a result of damage of reactors escaped into air iodine radioactive isotopes which were dispersed by air masses over Europe and Slovakia. Isotope I-131 was identified in samples of the atmosphere and the abstraction of Radiation Control SE EBO. The air from the atmosphere contaminated with isotopes of iodine from the Fukushima ventilation systems that do not contain iodine filters, sucked into the interior of the controlled area, then released in organised way and then measured in the ventilation chimneys of EBO NPP. The measured values thus entered a balance of radioactive discharges. Drain of I-131 from SE EBO was in that period plus a contribution coming from Fukushima NPP and measured activity I-131 had to be corrected.
Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp
Russell, Charles R
Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor
Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.
This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.
A modular system is suggested of monitoring the temperature field of the WWER reactor core. Standard measurements were completed with specified measurements and evaluation of coolant heating, with tests of functionality of measuring chains for coolant temperature at the fuel assembly outlet, changes in the offset of the thermocouple cold ends and with the evaluation of thermal and hydraulic characteristics of the core and of the primary circuit. Experience is presented with the application of this system based on the hardware and software of the RPP-16S control and computer system of the first and second units of the Bohunice V-1 nuclear power plant. Described are the structure of computer subsystems, the design of algorithms for data acquisition, testing of the credibility of temperatures measured with thermocouples, the processing of measured values and the algorithm for reactor heating computations. The system has been proven, and specified measurements made it possible to maintain a more stable and more accurate thermal output of the reactor. (M.D.)
This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)
Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference
There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted
Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)
The results of testing of equipment at Bohunice NPP and pipeline systems at Unit 3 of Kozloduy NPP (WWER-440 type reactors) are presented in this Final Report. These results side by side with experimental values of natural frequencies and decrements also include experimental data about vibration modes of tested equipment and pipelines. For the first time the results of new calculational-experimental examination of equipment seismic resistance at Unit 2 of Armenian NPP are presented. At Kozloduy NPP direction's request the planed additional tests of some selected items were put off on 1997. Instead of postponed tests we carried out detailed analysis of our past inspections of numerous equipment seismic resistance at the Unit 5 of Kozloduy NPP. Experimental data with results of additional analysis are presented
In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed
Scientific conference deals with problems in radiobiology, photobiology and radio-environmental sciences. Some papers deal with the historical aspects development of reactor accidents (Chernobyl NPP and NPP A-1 Jaslovske Bohunice) as well as history of nuclear sciences in former Czechoslovakia. Proceedings contain forty-seven papers
The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)
The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author)
Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric
[EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.
Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks
This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility
Bohunice NPPs fulfilled planned power supply to 103.4% in 1997 having supplied 9,968,758 MWh to the national power grid. Bohunice NPPs power generation reached 10,796,904 MWh which represented 51% share of Slovak power stations plc power generation and 44% of national power generation. Total Bohunice NPPs production represented 178,715,793 MWh of power since the start-up of their WWER 440 units. Together with the A-1 plant it represented 180,180,372 MWh in total since the Bohunice site had been commissioned. Heat supply for heating purposes to the town of Trnava was reliable and met customer's requirements representing the amount of 1,003,500 GJ of heat, which was by 221,173 GJ less than in 1996. The heat supply to Trnava decreased due to the heat line from 13 May to 3 October 1997, due to starting the heat line to the towns of Hlohovec and Leopoldov up, and due to warmer weather during the heating season. The heat supply to Trnava represented 8,787,323 GJ since the Bohunice-Trnava heat line had been started up (December 1997). The Heat Transmission Plant Jaslovske Bohunice caught 65,687 GJ of heat last year which represented 310,982 GJ since its start-up (November 1992). Total heat supply from nuclear sources for heating purposes was 1,691,052 GJ in 1997, which represented 16,495,333 GJ since the Transmission Plant had been started up. Performance indicators from the beginning of operation to the end of 1997 and performance indicators in 1997 as well as operation history of Bohunice 1 - 4 reactor are presented
Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date
A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.
Miller, H.I.; Smith, R.C.
This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.
This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)
Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...
A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)
In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)
The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components
A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.
In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel
Lasserre, T; Lasserre, Thierry; Sobel, Henry W.
We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.
This contribution deals with planning, management and organizational aspects of decommissioning of NPP shut down due to the accident (prototype NPP A1) and NPP shut down after normal operation (NPP V1). The A1 and V1 NPPs are located very close in Bohunice nuclear site however both plants have very different technology and operational history. The preparation of A1 NPP decommissioning strategy and relevant decommissioning plans was long term process, because the plant was shut down after the accident in 1977 and decommissioning was implemented first time in Slovakia with many specific difficulties. The decommissioning planning of V1 NPP was shorter and easier, because the plant was shut down after normal operation, there were lessons learned from the A1 NPP decommissioning planning, available legislation, available financing etc. Development of decommissioning strategies, preparation and planning for decommissioning, development of legislation for decommissioning, management of decommissioning projects and other aspects are described and compared. Lessons learned are formulated on the basis of analysis of past, ongoing and planned decommissioning activities in Slovakia. (author)
A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days
In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)
In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)
Report of the ASSET (Assessment of Safety Significant Events Team) follow-up mission to the Bohunice (units 1-2) nuclear power plant in Slovakia 5-9 July 1993. Root cause analysis of operational events with a view to enhancing the prevention of accidents
This Report of the IAEA Assessment of Safety Significant Events Team (ASSET) presents the results of the team's review of the status of implementation of the recommendations made by the 1988 ASSET mission to Bohunice nuclear power plant in Slovakia, and of progress made by plant management in prevention of incidents. The findings, conclusions and suggestions presented herein reflect the views of the ASSET experts. They are provided for consideration by the responsible Slovakian authorities. The ASSET team's views presented in this report are based on review of the documentation made available and on the discussions with plant staff. The report includes the official response of the operating and regulatory organizations of Slovakia to the ASSET findings and conclusions. Figs, tabs
A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)
The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)
Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures
A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.
The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)
A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)
In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)
The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.
A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance
It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)
The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated
Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)
Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)
A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.
Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.
A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.
Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)
Full text: The Atominstitut Vienna operates a 250 kW TRIGA Mark-II reactor since March 1962 used for nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection, dosimetry, low temperature physics and fusion research. During the past 20 years about 640 students graduated with a diploma - or PhD degree from the Atominstitut attached to the University of Technology Vienna. To perform nuclear relevant academic studies the Atominstitut offers about 100 highly specialised theoretical lectures and about 10 practical courses where students have to perform experiments in small groups of four on subjects mentioned above. Although the TRIGA reactor is a rather low power research reactor it is very easy and cheap to operate and an excellent tool to transfer knowledge and experience to the younger generation. This reactor is therefore not only used by other European universities such as University of Manchester or Bratislava Technical University but also by nuclear institutions such as the GRS/Germany, NPP Bohunice and NPP Mochovce for nuclear training. On an international scale the Atominstitut co-operates closely with the nearby located IAEA in international research projects, coordinated research programs (CRP) and supplying expert services. Regular training courses are carried out for the IAEA for Safeguard Trainees, fellowship places are offered for scientists from developing countries and staff members carry out expert missions to research centres in Africa, Asia and South America. In the past 20 years more than 120 IAEA fellows from all over the world have been trained at the Atominstitut. The fellows spend between one to twelve month at the Atominstitut and are integrated in the respective work program. Experience showed that out of this fellowship a long-term relation between the institutes continues. The paper focuses especially on the transfer of knowledge between
A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)
The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)
Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed
Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors
HbA1C test; Glycated hemoglobin test; Glycosylated hemoglobin test; Hemoglobin glycosylated test; Glycohemoglobin test ... have recently eaten does not affect the A1C test, so you do not need to fast to ...
University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators
This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.
After the accident at A-1 reactor in Jaslovske Bohunice NPP (Slovakia) large quantities of soil were also contaminated by cesium-137 after leaks from the liquid waste tanks. This soil was later excavated and is now stored in dedicated radioactive waste depots. Various decontamination methods were tested for the remediation of the samples of this soil. In this paper the electrochemical procedure, electrokinetic mobilisation and thermodesorption procedure were tested
The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor
Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)
Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)
This paper describes the current status of liquid waste (evaporator concentrates) inventory at V-1 and V-2 NPPs in Jaslovske Bohunice and the intention to separate boron from them with respect to waste minimisation and improvement of physical and chemical properties for further waste treatment and conditioning. Preliminary results of laboratory experiments concerned to borate crystallisation after pH adjustment with nitric or formic acid performed in the 1998 are given. At the present time laboratory experiments continuing - next acids, coagulation with carbon oxide, electrolytic process, ion exchange resin, study of decontamination factors, immobilization of boric acid, decrease radioactivity, purification of boron-contained compounds. Slovenske Elektrarne have accumulated 7,000 m3 of evaporator concentrates containing 100-180 g/l borate. In order to make more storage space available, it is proposed to remove some of the borate in the liquor by precipitation as sodium tetraborate and immobilise in either cement of bitumen. The supernate can be further volume reduced by evaporation and returned to the tanks. Slovenske Elektrarne are currently evaluating acid addition to the pH 12-13 concentrate to reduce the borate solubility. However, this adds to the salt burden of the waste through this chemical addition -thus creating future increases in conditioning and disposal costs. Boric acid is used in pressurized water as a soluble neutron poison to control reactivity and also to assure a safety margin in the spent fuel pool and during refuelling operations. Boric acid is also present in the water reserved for injection into the reactor in the event of postulated accidents. (author)
'Full text:' A 1-D model has been developed for testing different designs of hydride reactors. The computer program can simulate a complete reactor or a part of it in planar, cylindrical or spherical geometry. It reproduces an experimental loop: absorption followed by desorption and calculates heat transfer during the reaction. Simulation results have been confronted to experimental data with very good correlation. A reactor with a heat transfer matrix inside, such as aluminum foam, can be simulated. We have evaluated the size limits of a reactor and the category of foam that preserves the good reaction kinetic performances of a reactor filled with LaNi5. (author)
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
This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)
A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise
The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre
SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed
This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel
The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)
In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)
Presentation describes the history, present and future of the spent fuel management in the Slovak Republic. First experiences with spent fuel were gained in the seventies. Spent fuel form A-1 NPP was handled at Jaslovske Bohunice site, in order to prepare the spent fuel for the transport to the former USSR. After shut down of the A-1 NPP, all spent fuel was transported to the USSR. In 1978 first unit of V-1 NPP was set into operation. Actually there are six NPP units of the WWER type at Jaslovske Bohunice and Mochovce sites in operation in the Slovak Republic. These six units produce about 500 spent fuel assemblies per year. In 1988 an Interim spent fuel storage facility was build at Jaslovske Bohunice site. These facility stores spent fuel from four Jaslovske Bohunice units. In 2000 this facility was subject to a reconstruction, seismic upgrade and capacity enlargement. In 2004 Nuclear Regulatory Authority of the Slovak Republic approved transport container C-30 for transport of forty-eight spent fuel assemblies. The transport capacity has risen, so the number of transports could be reduced. In 2006 Slovak Electric Plc. (SE) will start transports of spent fuel from Mochovce site to Interim spent fuel storage facility (ISFSF) Jaslovske Bohunice. In addition, a project of Interim spent fuel storage facility at Mochovce site is going on. In the future Slovakia plans to find definitive solution for the spent fuel. One solution could be reprocessing and further usage in the power reactors, the other solution could be final deposition of spent fuel. (author)
Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.
A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.
The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)
A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)
The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)
The paper deals with experience and techniques in the application of remotely controlled robotic devices for the dismantling of the A-1 NPP technological equipment during undergoing decommissioning process of the A-1 NPP, which is characterized by high level of radioactivity and contamination. For liquidation of the heavy water evaporator has been applied a mobile robotic system MT 80, which had been developed, designed and constructed as a general-purpose decommissioning equipment. The heavy water evaporator as a part of the NPP heavy water system is located inside the main production unit building in Room No. 220 where the inner surface contamination is from 101 Bq/cm2 to the level of 103 Bq/cm2, dose rate up to 1.5 mGy/h and the feeding pipeline contained LRAW with high tritium content. The first step was the development of a work procedure with special focus on the elimination of activity and aerosols leaking into the environment. Special tooling was developed for application with the robot, such as hydraulic shears, circular saw, reciprocating saw, circular pipe cutter and a system for quick tool-change without direct intervention of the operators. Then, civil engineering modifications were made to the workplace and new technology was installed, including an efficient exhaust system. After draining-off the remains of LRAW and rinsing the pipeline, fragmentation of the pieces of equipment was started. The fragments are being deposited into drums which are transported in shielded containers to the decontamination facility, where their activity is reduced prior to their storage or further use. All operations are remotely controlled on basis of visual information from four cameras, with consistent radiation protection of the operators. This experience will be exploited in the continuation of work in Room No. 219, where the second identical heavy water evaporator is located. Nuclear Power Plant A-1 is situated in the locality of Jaslovske Bohunice. The A-1 NPP
The SCORPIO-VVER core monitoring system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic) and two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system. By both Czech and Slovak nuclear regulatory bodies it was licensed as a Technical Specification Surveillance tool. The monitoring system operates in two modes: in core follow mode and in predictive mode. In the core follow mode, the present core state is evaluated by a method combining the instrumentation signals and the theoretical calculation of the power distribution done by the core simulator. This procedure is followed by an automatic limit checking, where characteristics of the current state are compared to the Technical Specifications. The operator obtains relevant information on core status through the dedicated Man-Machine Interfaces. In the predictive mode, the operator can visualize the core characteristics during the transients forecasted for coming hours or days. Quick forecasts realized by the strategy generator are deeply analyzed by the predictive simulator. Similarly as in the core follow mode, characteristics of the evaluated states can be compared against Technical Specifications. Since it's first installation, the development of SCORPIO-VVER system continues along with the changes in WWER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The latest most significant changes were done in connection with implementation of a new digital I and C system, loading of the optimized Gd2 fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics) and improvements in the predictive part of the system (Strategy Generator). The currently finished
This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate
Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)
In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)
Stacey, Weston M
Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme
... to minimize the complications caused by chronically elevated glucose levels, such as progressive damage to body organs like the kidneys, eyes, cardiovascular system, and nerves. The A1c test result ...
While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident
MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)
With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements
Miller, Jan D; Hupka, Jan; Aranowski, Robert
A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.
The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel
The North Carolina State University (NCSU) PULSTAR reactor is a 1-MW light water pool-type reactor that began operation in 1972 as part of the university's land grant mission for teaching, research, and service. The nuclear services program was formed at the same time to develop and provide nuclear analytical services for members of the university research and industrial community. The majority of these services are neutron activation analysis (NAA) and low-level counting. Other services include neutron radiography, prompt gamma analysis, and neutron depth profiling. Industrial short courses on radiation safety and radioisotope techniques are offered regularly. The PULSTAR reactor facility has more than 800 visitors per year, most of whom are secondary school students participating in reactor-sharing activities
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
The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)
Stacey, Weston M.
An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.
Please note that the revised safety code A1 entitled 'MEDICAL CODE' is available on the web at the following url: https://edms.cern.ch/document/335476/last_released Paper copies can also be obtained from the SC Secretariat, e-mail : email@example.com SC Secretariat
The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)
This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)
The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially
The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs
The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)
Todd, Donald; Tsvetkov, Pavel
Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...
This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized
The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. 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. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab
As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described
Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy
The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported
This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)
The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported
Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)
Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)
Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)
The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost
The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)
Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)
Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant
In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)
INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety
The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)
A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling
The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed
The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)
Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)
The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)
Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)
Skaates, J. Michael
Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)
Van Walle, E
The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)
The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)
The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.
The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported
The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)
This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)