Sizewell 'B' power station - engineering and construction

International Nuclear Information System (INIS)

Walker, A.

1989-01-01

The Sizewell 'B' design is based on the Callaway PWR station in Missouri, USA. There have been many papers describing the additional criteria covering safety, operational experience and design and manufacturing practices which were applied to the Callaway reference design and which led to the Sizewell 'B' design concept. This paper describes the implementation of some of these design criteria leading to the layout related design features. Comparisons are drawn with the reference design. This paper concludes by describing various aspects of the design and construction process and the progress which has been made on the Sizewell site. (author)

Inspecting Sizewell B Class 1 and IoF components

International Nuclear Information System (INIS)

McNulty, T.

1990-01-01

Class 1 and ''incredibility of failure'' (IoF) pre-service inspections for Sizewell B cover steam generator and pressurizer welds and nozzles, welds on the emergency boration system, accumulators and reactor coolant loop, and the whole of the Class 1 piping. The contractor for this work has recently been selected. The scope of the work and the techniques and equipment to be used are outlined. (author)

Sizewell B

International Nuclear Information System (INIS)

1983-03-01

The probability of an aircraft crash on the proposed Sizewell B nuclear power station leading to a significant release of radioactivity is discussed. The aircraft considered are all types of civil and military fixed wing aircraft and helicopters. It is concluded that the general approach to the safety case for aircraft crash is acceptable. Outstanding issues are identified and a programme of additional work is discussed. (UK)

Radiation monitoring at Sizewell B PWR

International Nuclear Information System (INIS)

Hills, O.C.

1992-01-01

Radiation monitoring in Sizewell-B Power Station most significantly differs from that in existing UK Power Stations in two respects: firstly in the large number of on-line radiation monitors and secondly in the way that the monitors are linked into a fully-integrated, centralised data acquisition and display system, which can be accessed and viewed by the operators. An overview is given of how full process data is transmitted along data links between the Auxiliary Shut-Down Room, Technical Support Centre and Main Control Room, enabling Health Physics and other staff to access information from any radiation monitor. The permanently installed monitors together with the safety category, type and location are listed. As part of the Sizewell-B Process Plant control and instrumentation contract, NEI is to supply the Health Physics Instrumentation (HPI) and Process and Effluent Activity Monitoring System (PEAMS) (excluding the Primary Protection System) plus the Nuclear Sampling System (NSS). This paper concentrates on the HPI, and parts of the PEAMS and NSS for which NEI have the responsibility for system design, detail design, manufacture, site installation and commissioning. Section 2 briefly describes the sources of radiation at Sizewell-B; Sections 3, 4 and 5 describe the PEAMS, HPI and NSS respectively. Section 6 details the design of two of the Sizewell-B PEAMS subsystems. (Author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Harrison, J.R.

1982-11-01

An overview is given of the Safety Case for the Sizewell B nuclear power station as presented in the Pre-Construction Safety Report. Information which has been made available since its publication is included. Safety is considered for normal operation of the reactor and under fault conditions. Faults considered are those due to external hazards such as earthquakes, extreme weather conditions and aircraft crashes, internal hazards such as fire and missiles and reactor faults. Reactor fault studies described include transient analyses of pressurised faults and of loss-of-coolant accidents and the evaluation of the radiological consequences of design basis faults. (U.K.)

Description of Sizewell B nuclear power plant

International Nuclear Information System (INIS)

Meyer, G.; Stokke, E.

1997-09-01

The intention of this report is to present a condensed technical description of Sizewell B in a language understandable to non-technical personnel. It is unavoidable that some parts will be less precise than the technically initiated would like to see, but hopefully the content still give a realistic picture of Sizewell B. The technical description is based on publicly available material, of which the Sizewell B safety report has been particularly useful. Nearly all figures and drawings found in this description are reproductions of corresponding material in the safety report. To keep the description from becoming too voluminous it has been necessary to condense some background material down to a small volume. Hopefully this has not introduced any errors or inaccuracies, possible oversimplification at certain points must be weighed against the wish to cover most of the topics in the agreed table of contents for these NKS reports. (au)

Development of operator training programmes for Sizewell B

International Nuclear Information System (INIS)

Birnie, S.

1988-01-01

In accordance with existing practice, it is the Central Electricity Generating Board's (CEGB's) intention that the station manager of Sizewell B and future pressurized water reactor (PWR) stations will be responsible for ensuring that his staff perform their designated duties competently. To assist the station managers in fulfilling these responsibilities, the CEGB ensures that all nuclear training needs are identified, effective training strategies are developed and training programmes provided. The management, operation and development of the CEGB Nuclear Power Training Centre is integrated with nuclear power training activities conducted on site and at other locations. A systematic approach to training must be used so that the training is effective, i.e. that the staff can operate the plant safely and economically. To aid in the systematic production of PWR training programmes in general, but in particular for shift operations engineers, a PWR section was established in 1983 at the CEGB Nuclear Power Training Centre. A condition in the site licence for Sizewell B states that a suitable simulator must be available for training operations staff at least one year before fuel loading commences. The work of this section in operations engineer training is summarized. (author)

Electronic document management at Sizewell B

International Nuclear Information System (INIS)

Rippon, Simon.

1996-01-01

Sizewell ''B'', Britain's first PWR, officially opened on March 25 1996, will now rely for its document management on a sophisticated computer-based system. One of the largest single engineering projects ever to be commissioned on one site in Britain, Sizewell ''B'' accommodates more than 300,000 documents, including over 200,000 drawings. The electronic document management system will provide a number of important benefits, including a more direct method of maintaining the station's Configuration Management and hence maintaining high safety standards; improved turnaround in plant modification proposals (PMP); significant time and cost savings in managing vital records; and increased productivity. (Author)

Sizewell B: consent application for Britain's first PWR power station

International Nuclear Information System (INIS)

1981-02-01

The Central Electricity Generating Board has applied to the Secretary of State for Energy for consent and for other necessary permissions to construct a nuclear power station of about 1200 MW output capacity based on the pressurised water reactor (PWR) system on the Board's existing site at Sizewell (near Leiston) in Suffolk to be known as Sizewell B. Application has also been made to the Health and Safety Executive to extend the existing nuclear site licence to permit the use of the site for a pressurised water reactor. The Secretary of State for Energy has already stated that a Public Inquiry will be held into the application and this is expected to take place in 1982. The Board is making these applications now to give ample time for public discussion and consultation. Construction of the station could not begin until the outcome of the Public Inquiry is known and the necessary consents, nuclear licence and clearances have been given. The text of the application is presented. Some background information is given. (author)

The Sizewell B inquiry and East Anglia: the legacy of Three Mile Island

International Nuclear Information System (INIS)

Kemp, R.V.

1985-01-01

This paper argues that the great length and cost of the Public Inquiry into the proposed Sizewell-B PWR reactor is due to its investigatory style. This style became necessary because of the complexity of the issues, the difference in resources between the proponents and objectors and because of delays in licensing of the PWR by the Nuclear Installations Inspectorate. The fears about the safety of the PWR type reactors followed the accident at Three Mile Island in 1979, and the Major of Harrisburg was called by the local residents opposed to the siting to give evidence at the Inquiry. The ways in which the Sizewell-B Inquiry has attempted to come to terms with the fears and concerns of local objectors to the project are looked at. The paper considers how these can be properly weighed when the government is committed to nuclear power. The Inquiry is only advisory - the final decision will be taken by the Ministers after Parliamentary debate. The procedures and facilities adopted by the Inquiry are listed and, in particular, the two evening public meetings held in nearby towns to hear the views of local residents outside the formal main hearing are noted. The two main safety aspects of concern were the incidence of leukemia in East Suffolk, and the planning in case of an emergency at Sizewell-B. The experience of the Mayor of Harrisburg was relevant to this point. (U.K.)

The validation of corrosion inhibitors for ancillary cooling systems at Sizewell B PWR

International Nuclear Information System (INIS)

Allan, S.J.; Garnsey, R.; Lawson, W.F.

1994-01-01

Sizewell B has several low temperature ancillary cooling circuits for removing low grade heat. One of the principal circuits is the Component Cooling Water System (CCWS) primarily constructed from carbon steel pipework but also containing stainless steel, titanium, copper and cupronickel components. Linked to the CCWS is another carbon steel system namely the Reserve Ultimate Heat Sink (RUHS) a system unique to Sizewell B providing an independent heat removal capability for the CCWS as a safety back-up. Both systems utilize demineralized water with the addition of an inhibitor to prevent corrosion. The vendor recommended inhibitor namely potassium chromate is highly toxic to marine life and initial discussions with the regulatory authorities indicated that it would be extremely difficult to obtain a discharge consent to cover leakage or potential system purges and dumps associated with maintenance operations. Thus an alternative was sought. Sodium nitrite (anhydrous NaN0 2 )/borax (Na 2 B 4 0 7 1OH 2 0) inhibitor solutions have been used for many years in the power industry for preservation of carbon steel plant and several proprietary formulations are based on this mixture. There has been extensive test work in the UK on this inhibitor for secondary side boiler storage on gas reactors and therefore it was considered a prime candidate for use at Sizewell B. Starting in 1985 a programme of work sponsored by Nuclear Electric plc has been completed by NNC Ltd to compare nitrite/borax with other suitable inhibitors and validate its use for Sizewell B. (authors). 5 figs., 1 tab

Structural integrity of Sizewell B - the way forward

International Nuclear Information System (INIS)

Geraghty, J.E.

1996-01-01

The Incredibility of Failure (IOF) approach has been used on Sizewell B. This involves demonstrating that under all conditions, the failure of certain components is so remote as to be considered incredible. It means that the integrity of these components is such that they must remain intact assuming other plant items do fail. For example, a complete failure of the main loop pipework must be withstood by the RPV and steam generator shell. The IOF components are mainly those of the reactor coolant circuit and the purpose of this paper is to describe the approach adopted for the RPV, the steam generator shell, pressurizer and reactor coolant pump casing. The two arms of the safety case are achievement and demonstration of integrity. (author)

Sizewell B

International Nuclear Information System (INIS)

1983-07-01

The design strategy for keeping all exposures of persons to ionising radiations as low as is reasonably practicable (ALARP) is discussed for the proposed Sizewell B nuclear power station. The safety assessment objectives are discussed together with economic factors, choice of primary circuit materials to minimise the formation of crud, operational chemistry, filtration, decontamination, design modifications and radioactive waste management. It is concluded that a significant attempt has been made to quantify design decisions in respect of the ALARP strategy for dose reduction. A programme of additional work is discussed. (U.K.)

Sizewell 'B' power station public enquiry: CEGB statement of case. 2 vols and appendices A-G, H, J, L, M

International Nuclear Information System (INIS)

1982-04-01

This statement of case summarises the case which the CEGB intends to put to the public enquiry into building a PWR reactor at Sizewell. The two volumes are divided into chapters covering an introduction, the Electricity Supply Industry and the CEGB, nuclear development on the CEGB system, new generating plant, the need for Sizewell B and various aspects of safety and design. (U.K.)

Electrical and control aspects of the Sizewell B PWR

International Nuclear Information System (INIS)

1992-01-01

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

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

McFarlane, J.D.

1982-11-01

The NNC construction programme for the Sizewell B nuclear power station is outlined. The following topics are considered: constructability aspects of the Sizewell B design; overall project programme and organization; NNC target construction programme; construction planning; material quantities and installation rates; site construction manpowers; construction contracting policy; site industrial relations and productivity; site management. (U.K.)

Sizewell 'B' public inquiry

International Nuclear Information System (INIS)

Barritt, E.E.

1982-11-01

The report falls under the headings: introduction (need/economics; safety; waste management; local and site specific issues); the Sizewell site - history and planning policies; planning and environmental implications; land use and landscape implications; ecological implications; mineral requirements; highway implications; employment implications; accommodation implications; infrastructure and social/community services; safeguarding restrictions for Sizewell - Nuclear Installations Inspectorate; decommissioning. (U.K.)

Sizewell B PWR: safety implications for operating staff. A report

Energy Technology Data Exchange (ETDEWEB)

1983-01-01

A report given on the safety implications for the staff who would be involved in the commissioning and operating of Sizewell B reactor, looking in particular detail at the following aspects of the plant and its proposed operation: operator access to the containment whilst the reactor is on-load and the reasons for and means of restricting this, the use of robotics to minimise routine access to high radiation areas, circuit chemistry in relation to its effect on minimising the coolant activity, the handling and storage of the radioactive waste arisings on-site, including the use of robotics and the integrity of the pressure vessel as considered by the Cottrell/Marshall dialogue.

Reactor choice - the merits of Sizewell follow-up and longer-term options

International Nuclear Information System (INIS)

Smith, P.

1994-01-01

During 1994 Sizewell B will be completed, a contractor will be selected for the Lungmen Project in Taiwan and the Nuclear Review will hopefully clear the way for continuing PWR construction in the UK. Once the Government has sanctioned the next nuclear station, the choice of plant will fall to Nuclear Electric. Consideration of the available options demonstrates clearly that for the short and medium term only Sizewell C, a twin replica of Sizewell B, provides the low-risk option essential to attracting private finance. The UK has built up the design, manufacturing and construction capability necessary to build the Sizewell C design, both in the UK and overseas. However, there is only a limited window of opportunity to exploit these capabilities since they will decay unless sustained by continuing orders. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Jenkin, F.P.

1982-11-01

The CEGB's planning and assessment procedures relating to Sizewell B are explained. The various bases of need for new generating plant capacity are outlined. These are: capacity requirement; provision of economic generating capacity; security of fuel supplies; making provision for future investment options. A detailed case for Sizewell B is then presented, including an assessment of risk and uncertainty. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Edmondson, B.

1982-11-01

A discussion is given on the safety issues for the pressure circuit components of the proposed Sizewell B nuclear power station. The detail of the evidence concentrates on the reactor pressure vessel itself, as the principal example of a component whose failure is considered incredible. The general background to construction and operation of pressurized plant is described. A description of the form of the vessel and its role is given. Each of the factors which lead to confidence in the integrity of the vessel is then considered. (U.K.)

The structural integrity safety case for Sizewell B power station

International Nuclear Information System (INIS)

Gerachty, J.E.

1993-01-01

This paper presents the safety case approach adopted for the components of the Sizewell 'B' Power Station for which a high degree of structural integrity is required. Such components include the Reactor Pressure Vessel, Steam Generator and Pressuriser for which Incredibility of Failure is claimed. The two parts of the case involve achievement and demonstration of integrity. This is achieved by extensive measures involving design, manufacture, materials and inspection. The demonstration has required a fracture mechanics approach. The specific role of inspection validation and its relation to critical defect size is described. (author)

The design and management of cables and supporting steelwork systems for Sizewell 'B'

International Nuclear Information System (INIS)

Kerry, L.P.

1992-01-01

Sizewell B Power Station has over 33,000 cables connecting in excess of 20,000 items of equipment. Over 120 cable types are used. Many cables supply power to, or monitor, safety category 1 equipment essential to the safe operation of the Reactor and supporting systems. Cables must be designed to survive the postulated hazards of Reactor systems fault, earthquake, fire and radiation exposure. The cable network is the nerve system of the station; its role is to distribute electrical power and control indication, metering and protection signals to the various items of plant and equipment under normal and emergency operating conditions. To perform these functions satisfactorily a new range of cable insulations had to be developed and supporting steelwork designed to withstand the effect of an earthquake. The 63 month Sizewell B construction programme means record cable laying and completion rates have to be achieved needing careful and detailed planning from design to installation. As with electrical equipment, cables are designated either essential (to Reactor safety operation) or non-essential. This paper describes the required safety design criteria, how these are achieved and the design approaches to seismically qualified supporting steelwork. Also described are the cable design, scheduling, routing and planning processes which lead to the required installation and commissioning activities being completed in a controlled and timely manner. (Author)

Diaphragm walling for Sizewell B sets records

International Nuclear Information System (INIS)

Anon.

1988-01-01

The first phase of construction of the Sizewell-B nuclear reactor has been completed. This was the building of a diaphragm wall around the site. It is one of the largest and deepest diaphragm walls to be installed in Europe. The site can be pumped dry of groundwater and the foundations constructed in the dry. The specifications of the wall and its construction, using two Hydrofraise excavation rigs, are described. The excavated material is brought up as a slurry and the (bentonite) slurry is cleaned and desanded. Most of the wall has been formed using a plastic concrete but reinforced concrete has been used for some stretches. The diaphragm wall, which is 1258m long and 55m deep on average, was built in 19 weeks. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Newell, J.E.

1982-11-01

A brief description is given of the design of the steam generators for the proposed Sizewell B power station. A review of the operational behaviour of earlier models is presented. The problem of degradation of steam generator tubes is considered in some detail. Model F steam generator, which has been incorporated into the Sizewell B reference design, is discussed. Finally the impact of steam generator tube degradation upon radiological safety is discussed. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Passant, F.H.

1982-11-01

The radioactive wastes expected to arise at the proposed Sizewell B PWR are identified. The plant provided to collect, process, store and dispose of the wastes is described. The quantities and activities of wastes for disposal are derived. The proposed methods of transport and disposal are described. Wastes which would arise from reprocessing and from decommissioning Sizewell B are discussed. Finally some of the radiological aspects of waste disposal are outlined. (U.K.)

Ultimate load model test for Sizewell 'B' primary containment

International Nuclear Information System (INIS)

Crowder, R.

1988-01-01

This paper considers the factors influencing the adoption of an ultimate load factor for the Sizewell 'B' PWR primary containment structure. As part of the validation process for the ultimate load analysis method, a proposal has been made by Nuclear Design Associates to build and test a 1/10th scale model of the containment structure, which would proceed following the granting of section 2 consent for Sizewell 'B'. The modelling principles, construction method and test proposals are examined in some detail. The proposal is currently being considered by the CEGB's Project Management Team. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Vignes, A.

1983-02-01

Evidence is presented to show how the integrity of the reactor pressure vessel to be constructed for the proposed Sizewell B nuclear power station could be achieved by the French company Framatome. Following a description of the organization and experience of Framatome, their ability to manufacture the pressure vessel to the preferred design and to meet all the requirements of the National Nuclear Corporation is demonstrated. Then it is shown, using examples, how the various procurement and fabrication operations are carried out and controlled and how the quality of the materials and products is assessed and guaranteed. (U.K.)

Sizewell and the people: what the community thinks

International Nuclear Information System (INIS)

Chappell, H.

1983-01-01

Some 15 or so of the people who live in Leiston, the town (pop. 4000) nearest the proposed Sizewell-B site, give their views about the building of the PWR station. The points raised include the possibility of employment, the disruption to the town during the actual building of the reactor and the safety record of PWRs. It is an issue which divides families and where young and old are found on both sides of the argument. The public inquiry is felt to be a formality only and the building of Sizewell-B to be inevitable. (U.K.)

Proof of evidence by National Nuclear Corporation Limited for Sizewell 'B' public inquiry

International Nuclear Information System (INIS)

Franklin, N.L.

1983-02-01

The history, experience and capability of National Nuclear Corporation Limited (NNC) and the requirements for the Sizewell B PWR project are reviewed. A Joint Project Team comprising staff from NNC, Westinghouse, Bechtel and the CEGB has been formed to develop the design of the station and the safety case. It is concluded that Sizewell B will be economic in its own right and will equip NNC with expertise, experience and resources in PWR technology. The construction of Sizewell B will provide the experience on which future PWR projects can proceed with confidence. (U.K.)

The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

International Nuclear Information System (INIS)

Hall, P.; Hutt, P.

1994-01-01

This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

Validation gets underway on Sizewell ''Incredibility of Failure'' components

International Nuclear Information System (INIS)

Anon.

1992-01-01

The Inspection Validation Centre (IVC) of AEA Reactor Services in the UK has begun an eighteen month programme to validate the procedures and personnel of OIS plc, the inspection agents chosen by Nuclear Electric to carry out the pre-service ultrasonic inspection of the Sizewell B Pressurized Water Reactor components assigned to the ''Incredibility of Failure'' (IoF) category. The work involves several Sizewell B primary circuit components - the steam generators, pressurizer, and primary pumps - and will consider the inspections to be applied to the circumferential and nozzle-to-shell welds, nozzle inner radii and the pump fly-wheel forging. The validation will provide independent confirmation that OIS personnel are capable of using manual and automated methods to find and size any flaws of structural concern in these components. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

George, B.V.

1982-11-01

The design of the proposed Sizewell B power station and the policies which have guided design choices are outlined. The specification for Sizewell B and the production of its reference design are discussed. The Nuclear Steam Supply System and the associated safeguards systems are dealt with in more detail. The protection against hazards causing damage to plant and the protection provided to reduce radiation exposure are described. Control, instrumentation and operation of the plant are discussed. A section on quality of design follows. Finally economic aspects of the design are dealt with. (U.K.)

Can the myth be maintained? - understanding the economics of Sizewell B

International Nuclear Information System (INIS)

Harper, M.

1990-01-01

The cost of electricity generated by Sizewell-B is discussed as this has been the model for the economic arguments in favour of the proposed Hinkley Point C pressurised water station. However, since the Energy Select Committee report concluding that electricity from Sizewell-B will be substantially dearer than from coal-fired stations was published public justification is now expressed in terms of benefits to the environment. Reduction of the greenhouse effect, security of supply and maintenance of the nuclear option are now more important. The economic case is examined. (UK)

A local authority view of the Sizewell B decision and the role of the public enquiry

International Nuclear Information System (INIS)

Ayre, D.

1987-01-01

The background to the Sizewell-B public inquiry is presented. The local authorities are important in this as it was they who first received the Central Electricity Generating Board's request to build Sizewell-B and who triggered the need for a Public Inquiry. The local authorities did not object in principle to the use of the Sizewell site for additional nuclear power generation provided the Government was satisfied that there was a national need for the extra capacity. In future inquiries there might not be acceptance in principle which would change the nature of the Inquiry. The local authority involvement in the public inquiry is discussed under the headings: need and economics, safety and local environment issues. The relevant parts of the Sizewell-B report are considered. The Secretary of State's decision, which follows the Inspector's report, justifies the confidence put in the inquiry process by the local authorities. Concern is expressed, however, that extra financial help is needed to pay for the local authorities expenditure on the Inquiry and also for the consequences of the Sizewell-B project itself. (U.K.)

Sizewell 'dirty tricks'

International Nuclear Information System (INIS)

Martin, Steve.

1987-01-01

The pro-nuclear lobby is accused of dubious tactics to promote their case for the Sizewell-B reactor. These include reassessing future electricity demand which, it is claimed, could only be met by nuclear power, claiming that new power stations were needed to avoid blackouts during cold spells, and the reporting of a major design fault in the control rods of the Torness and Heysham AGR stations. The latter is felt to be related to the promotion by the South of Scotland Electricity Board of the AGR case, as opposed to the Central Electricity Generating Board's advocation of a PWR reactor design. The author argues for a series of coal-fired power stations instead, and a major energy conservation programme. (UK)

Sizewell B nuclear power station: the basis for the decision by the Health and Safety Executive to grant consent to load fuel into the reactor

International Nuclear Information System (INIS)

1994-01-01

The licensing and consent process and the basis for granting a consent for Nuclear Electric to load fuel into the Sizewell B reactor in the United Kingdom are explained. Consent was granted by the UK Nuclear Installations Inspectorate on behalf of the Health and Safety Executive on satisfactory completion of construction and those commissioning stages needed to proceed safely, and the production of a satisfactory safety case. A summary of the assessment of the safety case is appended. It covers the reactor core, coolant system structural integrity, engineered safety features, main and essential electrical system, control and instrumentation, radioactive waste management, radiological protection, fuel storage and handling, civil works and structures, fault analysis, human factors, hazard analysis, quality assurance, and decommissioning. (UK)

The UK nuclear programme: The Sizewell experience

International Nuclear Information System (INIS)

Salter, W.B.

1990-01-01

The current status of the Sizewell 'B' PWR programme and the effect on it of the proposed privatisation of U.K electricity generation is reviewed. Departures from and additions to the Standard Nuclear Unit Power Plant System (SNUPPS) reference plant design are given. These include Reactor Coolant System overpressure protection and the addition of an Emergency Charging System and an Emergency Boration System. Improvements in monitoring Reactor Coolant System water level during refuelling and maintenance shutdown operations are presented. (author)

The economic consequences of the Sizewell 'B' nuclear power station

International Nuclear Information System (INIS)

Fothergill, S.; Gudgin, G.; Mason, N.

1983-01-01

The subject is covered in chapters, entitled: introduction (the background to Sizewell 'B'); policy options (Sizewell 'B'; a new coal-fired station; the no-new-station option; a PWR programme); economic framework (direct effects; financing; final macroeconomic effects); the construction phase (capital costs; direct effects; final effects; summary); the operating phase (a new power station as a replacement for older stations; the period of base-load operation; the later years of operation; summary); conclusions and policy recommendations. The first recommendation is that if a new power station is built it should be a coal-fired station rather than a PWR. The second recommendation is that if a new coal station is built there is a case for building it early, ahead of demand. (U.K.)

Ergonomics and its application to Sizewell 'B'

International Nuclear Information System (INIS)

Singleton, W.T.

1986-01-01

The scope of the ergonomics contribution to the design and operation of power stations is described on the basis of current experience in the CEGB and in other countries. The ergonomics questions which arise in relation to Sizewell 'B' are enumerated in detail. Issues which arise from the point of view of station operation and from the complementary point of view of human behaviour are considered. (author)

Sizewell 'B' PWR reference design

International Nuclear Information System (INIS)

1982-04-01

The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Henderson, B.

1982-11-01

Architectural concepts and design proposals associated with the proposed Sizewell B power station are outlined. The figures are in a separate volume. They consist of the site layout plan, an axonometric drawing of the site, an elevations drawing and a colour perspective drawing of 'A', 'B' and 'C' stations. (U.K.)

Is Sizewell a good investment

International Nuclear Information System (INIS)

MacKerron, G.

1984-01-01

The CEGB proposal, at present the subject of a public inquiry, to build a Westinghouse pressurised water reactor at Sizewell, is discussed. The principal aim of the paper is to examine the case for achieving cost savings for the electricity supply industry. The headings are: system cost savings (discussion of method of assessing savings); the CEGB case (involving various assumptions about the UK economy); objections to the Board's case (assumptions about fossil fuel prices, construction costs, operating characteristics and net present costs; uncertainty); other arguments for Sizewell (capacity need; fuel diversity); conclusion. (U.K.)

Designing power transformers for Sizewell 'B' as quality products

International Nuclear Information System (INIS)

White, A.; Poole, C.W.

1992-01-01

At the beginning of 1990, Nuclear Electric plc., then a Division of the CEGB, placed a contract with GEC ALSTHOM Transformers Limited of Stafford for the design, supply and installation of the main transformers for the Sizewell B electrical system. The following transformers are being provided under this contract; two Generator Transformers, two Station Transformers, two Unit Transformers and two Fault Limiting Reactors. The main objective of this paper is to describe the quality assurance philosophies applied to the design of these transformers to ensure that the performance requirements were met in full and to describe how these philosophies were converted into a practical working system. The paper does not set out to present a detailed technical appreciation of the transformers themselves but rather, to demonstrate that by addressing quality in the design and design processes, reliable and efficient items of high value, capital equipment result from the partnership in quality between supplier and customer. (Author)

ECOS: a configurable, multi-terabyte database supporting engineering and technical computing at Sizewell B

International Nuclear Information System (INIS)

Binns, F.; Fish, A.

1992-01-01

One of the three main classes of computing support systems is concerned with the technical and engineering aspects of Sizewell-B power station. These aspects are primarily concerned with engineering means to optimise plant use to maximise power output by increasing availability and efficiency. At Sizewell-B the Engineering Computer system (ECOS) will provide the necessary support facilities, and is described. ECOS is being used by the station commissioning team and for monitoring the state of some plant already in service. (Author)

Torness and Hunterston - as uneconomic as Sizewell B

International Nuclear Information System (INIS)

Jeffery, J.W.

1986-01-01

The admission by CEGB that Sizewell B would be uneconomic under conditions postulated in the evidence given by the present author at the Sizewell Inquiry, is used to show that Torness and Hunterston AGRs have almost no chance of being economic, because the large immediate increases in postulated real coal prices required to enable Torness to produce a lifetime surplus, have not materialized. The attempt of SSEB to conceal the fact that it postulated this large increase in real coal prices in order to make Torness appear to be economic, is described and exposed. The joint practice of CEGB and SSEB of discounting long-term nuclear costs so that they make no significant contribution to the final calculated unit cost is described. Finally, the recent document, explaining the SSEB's case for keeping the AGR, is analysed and characterized as 'one of the most misleading documents on the economics of nuclear power to be seriously presented to the public'. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Youngman, P.

1982-11-01

The scenic effect of the proposed Sizewell B power station is considered. Offsite planting of trees, landscaping of the areas around the new access road and of the site itself are discussed. The landscape designs are in a separate volume. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

McInerny, P.T.

1982-11-01

A description is given of the policy and practices that would be adopted in the commissioning, operation and maintenance of the proposed Sizewell B PWR power station. The system of personnel recruitment and training required to staff the station is discussed. (U.K.)

Software change control in the Sizewell B ISCO

International Nuclear Information System (INIS)

Johnson, A.

1997-01-01

Central to the control and instrumentation system of the Sizewell B nuclear power plant is a control, data acquisition and control system based on a distributed network of several hundred microprocessors. The system has been integrated into a single functional unity, and software modifications affecting one part of it have an effect on the other parts. The software modification and configuration processes are therefore kept as similar as possible. (A.K.)

Man-machine interface systems for the Sizewell B Nuclear Power Station

International Nuclear Information System (INIS)

Boettcher, D.B.

2004-01-01

Sizewell B is the first nuclear power station to be built in the United Kingdom using the Pressurised Water Reactor or PWR system. The design is based on stations operating in the United States, but many changes and new features have been introduced to bring it up to date, and to meet United Kingdom practice and regulatory requirements. The Man-Machine Interfaces (MMIs) in the control rooms have been newly designed from first principles, with special attention paid to human factors and the role of the operators. The instrumentation and control (1 and C) systems which interface the MMIs to the process plant, and automate the operation of the station, use advanced technology to achieve high performance and availability. This paper describes the development of the control rooms and 1 and C systems, explaining the thinking that lay behind the principal decisions. (author)

An exploratory study of the local socio-economic effects of Sizewell B

International Nuclear Information System (INIS)

1982-12-01

This Sizewell B study involved five main tasks: (i) to provide broad estimates of the temporary and permanent accommodation requirements of the proposed Sizewell B workforce and families and the likely level of local supply of accommodation; (ii) to provide broad estimates of the requirements for school places of the accompanied workforce and the likely local supply of such places; (iii) to assess further the wider employment effects of the proposed development on local firms, particularly as potential suppliers of goods and services; (iv) to conduct further studies of the expenditure patterns of power station employees; and (v) to incorporate where relevant the likely effects of a policy of double day shift work (in power station construction) into the predictions. Results of the study are reported. (author)

Probabilistic safety analysis for Sizewell B - proof of evidence, established on behalf of the Friends of the Earth, London, within the framework of the Sizewell B power station public inquiry

International Nuclear Information System (INIS)

Hahn, L.

1985-01-01

The conclusions drawn in the study presented by the Oeko-Institut, Freiburg, state that there are indeed advantages in the design of the PWR-type Sizewell-B reactor over other types with regard to certain event sequences, but that on the whole, with regard to overall accident probability, there is no significant progress detectable, or even provable. The main results of the analysis in hand undoubtedly are that the unrealistically favourable data presented in the other two Sizewell studies are brought about not by the design features, but by the assumptions, methods and data applied in the studies, so that certain weak points in the methodological approach common to both studies inevitably lead to similar results. The authors of the study in hand explain some of the weak points of the Westinghouse Study and of the CEGB Study to be as follows: - Lack of error bands of results. - Lack of a quantitative analysis of in-plant fire accidents. - Lack of quantitative analysis of external impacts. - Lack of (CEGB) or unsatisfactory (Westinghouse) evaluation of the measuring, control, and drive systems. - Lack of (CEGB) or unsatisfatory (Westinghouse) evaluation of the human factors. - Completely unsatisfactory evaluation of common-mode failures (i.e. multiple failures due to common cause, undiscovered systems interactions, etc.). - Unacceptable handling and use of available data (using data assessments made by experts, instead of data available from experience). (orig./HSCH) [de

Main unit electrical protection at Sizewell 'B' power station

International Nuclear Information System (INIS)

Fischer, A.; Keates, T.

1992-01-01

For any power station, reliable electrical protection of the main generating units (generators plus generator transformers) has important commercial implications. Spurious trips cause loss of generation and consequent loss of revenue, while failure to rapidly isolate a fault leads to unnecessary damage and again, loss of generation and revenue. While these conditions apply equally to Sizewell B there are additional factors to be taken into consideration. A spurious trip of a main generating unit may lead to a trip of the reactor with an associated challenge to the shutdown and core cooling plant. The generator transformers, besides exporting power from the generators to the 400 kV National Grid, also import power from the Grid to the 11 kV Main Electrical System, which in turn is the preferred source of supply to the Essential Electrical System. The Main Unit Protection is designed to clear generator faults leaving this off-site power route intact. Hence failure to operate correctly could affect the integrity of the Essential Electrical Supplies. (Author)

Radiation protection in the design and operation of Sizewell B

International Nuclear Information System (INIS)

Oldfield, M.

1989-01-01

This paper raises Health Physics operational issues specific to PWR or not previously encountered by the CEGB, which need to be resolved before the operation of Sizewell B and its replica successors. Staffing, access to contamination areas, training, temporary scaffolding and temporary lead shielding are discussed and the author makes some observations and suggestions. (author)

The use of probabilistic safety analysis in design and operation -- Lessons learned from Sizewell B. Annex 14

International Nuclear Information System (INIS)

Buttery, N.E.

2002-01-01

Probabilistic Safety Assessments (PSAs) have been used extensively in the design and licensing of Sizewell B. This paper outlines the role of PSA in the UK licensing process and describes how it has been applied to Sizewell B during both the pre-construction and pre-operational phases. From this experience a 'Living PSA' has been formulated which continues be used to support operation. The application of PSA to Sizewell B has demonstrated that it is a powerful tool with potential for future use. Its strengths and limitations as a tool need to recognised by both users and regulators. It is not a fully mechanistic means of ensuring design safety, but is an important aid to decision making. It also has the potential to allow risk judgements to be taken in conjunction with commercial and environmental issues. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Gittus, J.

1982-11-01

In volume one an analysis is presented of degraded core accidents involving the proposed Sizewell B PWR. Consideration is given to their estimated frequency, the course which they take, the release of radioisotopes from the plant and the radiological consequences of such a release. Volume two contains all the figures, tables and appendices presented as evidence. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Wilson, A.

1982-11-01

The methods of establishing construction time, capital cost, availability and lifetime of different types of generating plant are described. In volume one assessments are presented for a new coal fired station, for a new AGR station and for a new PWR station - Sizewell B. Volume two, contains all diagrams, tables and appendices presented as evidence. (U.K.)

The management of interfaces between organizations involved with the construction of Sizewell 'B'

International Nuclear Information System (INIS)

Butler, P.H.; Salmons, B.H.

1990-01-01

The overall organisation for the design, construction and commissioning of Sizewell 'B' is described, together with the specific responsibilities of the organisations involved. The paper describes how the parameters for each interface are identified, and how the formal arrangements for controlling transmission of information and material are handled. Whilst each organisation will perceive quality achievement as the ultimate objective, the specific individual responsibilities may dictate different, and sometimes opposing mechanisms for implementation, due to different perceptions of priority. The PWR Project Group has the ultimate responsibility for building Sizewell 'B' to cost and schedule and to the specified quality, and the paper describes, with some examples, how the different perceptions are addressed in order to achieve the ultimate objective. (author)

Industrial relations and site management proposals for Sizewell 'B'

International Nuclear Information System (INIS)

Burbridge, R.N.

1986-01-01

The vendor assessment and general contract strategy for the proposed Sizewell 'B' PWR are reviewed with particular reference to a 'Key Date' procedure. The family of programmes used in constructing CEGB projects is indicated and the intended site labour strategy for the project is outlined. To exemplify the current success of these policies, the paper concludes with a brief review of their application to the CEGB's Drax Completion Project. (author)

Remote filter handling machine for Sizewell B

International Nuclear Information System (INIS)

Barker, D.

1993-01-01

Two Filter Handling machines (FHM) have been supplied to Nuclear Electric plc for use at Sizewell B Power Station. These machines have been designed and built following ALARP principles with the functional objective being to remove radioactive filter cartridges from a filter housing and replace them with clean filter cartridges. Operation of the machine is achieved by the prompt of each distinct task via an industrial computer or the prompt of a full cycle using the automatic mode. The design of the machine features many aspects demonstrating ALARP while keeping the machine simple, robust and easy to maintain. (author)

Design and safety of the Sizewell pressurized water reactor

International Nuclear Information System (INIS)

Marshall, W.

1983-01-01

The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

The future of nuclear power after Sizewell B. 3 v.: v. 1 Economic issues; v. 2 Environmental and safety issues; v. 3 Public perception issues

International Nuclear Information System (INIS)

1987-01-01

The three days of conference proceedings are published in three separate volumes. The first includes 7 papers relating to economic issues - those presented at the Sizewell-B public inquiry and the changes in the economic situation since the inquiry ended. The electricity demand, how this demand is to be met by nuclear and other fuel sources and how energy conservation might be an economic alternative to simply building more generating capacity are all issues discussed. The possible privatisation of the industry is also touched on. Volume two has 8 papers concerned with environmental and safety issues. These include the influence of the Sizewell-B decision on nuclear licensing and reactor safety, the technical and safety aspects of pressurized water reactors (PWR), the roles of British Nuclear Fuels and the United Kingdom Atomic Energy Authority, and radiation protection and effluent discharge control. The six papers in volume 3 look at public perception issues - not only towards nuclear power but towards the public inquiry process. The local authority view, the Friends of the Earth case against the PWR, and technical expertise in the decision process are also topics covered. All the papers are indexed separately. (UK)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Gammon, K.M.

1982-11-01

The site selection and site specific aspects of the proposal to build Sizewell B PWR are dealt with. A description of the site and its surroundings is followed by physical and technical details of the proposed development. Consultations with statutory authorities, environmental and other local issues are discussed. Finally population distribution around the site and the way in which this could be controlled are considered. (U.K.)

Validation results of the pre-service ultrasonic inspections of the Sizewell B pressurizer and steam generators and reactor coolant pump flywheels

International Nuclear Information System (INIS)

Conroy, P.J.; Leyland, K.S.

1995-01-01

In the UK, concern over the safety issues associated with nuclear power generation resulted in a demand for a public inquiry into the construction and operation of Sizewell ''B'', Britain's first PWR. This public inquiry was additional to the UK's normal licensing process. The onus was placed upon the UK utility, CEGB (now Nuclear Electric plc) to provide evidence to the inquiry to support the case that the plant would be constructed and operated to a sufficiently high standard of safety. Part of the evidence to the inquiry (1) relied upon the ability of ultrasonic inspections to verify that the reactor pressure vessel and other safety critical components (collectively known as IoF components), were free from defects that could threaten structural integrity. At that time, the body of evidence showed that although ultrasonic inspection had the potential to satisfy this requirement, it would be necessary to validate the procedures and key operators used in order to provide assurance that they were adequate. Inspection validation therefore became an integral part of the UK PWR nuclear power program

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Dutton, L.M.

1982-11-01

A description is given of the main sources of radioactivity and the barriers which would prevent their release at the proposed Sizewell B nuclear power station. The methods of calculating the radiological consequences associated with the most significant faults within the design are outlined. The following possibilities are considered: faults involving high fuel temperatures; loss of coolant accidents; secondary coolant circuit faults; waste processing system faults; irradiated fuel faults. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Hawes, F.B.

1982-11-01

The biological consequences of using a direct cooled system at the proposed Sizewell B power station are dealt with. Problems caused by the impingement of organisms on the fine-mesh screens through which the water is pumped, by the entrainment of smaller organisms in the flow through the cooling system and by the discharge of warmed chlorinated water into the sea are discussed. The chlorination of cooling water is described. (U.K.)

Sizewell 'B' public inquiry. Proof of evidence on local environmental issues

Energy Technology Data Exchange (ETDEWEB)

Barritt, E E

1982-11-01

The report falls under the headings: introduction (need/economics; safety; waste management; local and site specific issues); the Sizewell site - history and planning policies; planning and environmental implications; land use and landscape implications; ecological implications; mineral requirements; highway implications; employment implications; accommodation implications; infrastructure and social/community services; safeguarding restrictions for Sizewell - Nuclear Installations Inspectorate; decommissioning.

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Wright, J.K.

1982-11-01

A general description of the nuclear fuel cycle for PWRs and AGRs is given. The details of the fuel cycle for Sizewell B are explained and its cost is analysed against a specific set of background assumptions. Then, in less detail, an evaluation of PWR fuel cycle costs is presented against a wide range of background assumptions. The cost of the AGR fuel cycle is analysed. Finally the uncertainties in the results obtained are considered. (U.K.)

Why mini Sizewell is safe as houses

International Nuclear Information System (INIS)

Pease, J.

1989-01-01

A one-tenth scale model of the Sizewell-B nuclear power station pressure vessel is under construction. The framework was made from steel, timber and lathe and plaster work. Reinforcement and prestressing can then be fitted. The object is to build a structure that will reflect the behaviour of the real thing. Once completed the vessel will be filled with water (not gas as in the real reactor) and the pressure brought up to 15% overload. 600 measuring instruments will make measurements of the mechanical properties. The model will then be tested to failure. The model will be analysed by computer, then tested, and the results compared. (author)

Assessment of options for the treatment of Sizewell PWR liquid effluent

International Nuclear Information System (INIS)

Hornby, J.; Allam, J.; Knibbs, R.H.

1992-01-01

This report describes the origins of PWR liquid waste streams, their composition and rates of arising. Data has been collected from operational PWRs and estimates obtained for Sizewell B PWR liquid waste streams. Current liquid waste treatment practices are reviewed and assessments made of established and novel treatment techniques which could be applicable to Sizewell B. A short list of treatment options is given and recommendations are made relating to established treatment technologies suitable for Sizewell B and also to development work on more novel treatments which could lead to a reduction in waste disposal volumes. (author)

A Review of the United Kingdom Fast Reactor Programme, March 1987

International Nuclear Information System (INIS)

Bramman, J.I.; Wheeler, R.C.

1987-01-01

Nuclear power produced about 20% of the electricity supply in the United Kingdom in 1986, mostly from gas-cooled reactors, i.e. the 10 AGRs currently in operation and the 26 older MAGNOX reactors. Plans to increase the nuclear component of generating capacity by building the first PWR in the UK, Sizewell 'B', were strongly endorsed in the report by Sir Frank Layfield published on 26 January 1987. This resulted from the Public Inquiry into building Sizewell 'B', which was held between 11 January 1983 and 7 March 1985, the longest Public Inquiry ever held in the UK. The government gave the go-ahead for the building of Sizewell 'B' on 12 March 1987

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Healey, T.; Board, S.J.

1982-11-01

The issue of fuel clad ballooning during a postulated large loss of coolant accident is examined. A summary of the evidence which shows that such an event in the proposed Sizewell B nuclear power plant would not lead to violation of the fuel limits laid down in the United States NRC Code of Federal Regulations, 10 CFR 50, Appendix K. The present status of the computer codes BART and TAPSWEL and the validation evidence to demonstrate their adequacy is outlined. (U.K.)

Sizing up Sizewell

International Nuclear Information System (INIS)

Anon.

1983-01-01

The CEGB proposal to build a PWR nuclear power plant at Sizewell, in Suffolk, is discussed under the headings: paying for power (economics of Sizewell); paying for capital (high capital cost of nuclear power plants); paying for delays (effects of construction delays and design changes to meet safety requirements); paying for coal (circumstances in which a fossil-fuel power plant would be less costly than expected); playing for time (advantage argued for postponing Sizewell's construction while pressing on with an energy conservation programme). (U.K.)

Electricity consumers' council statement of case to the Sizewell 'B' power station public inquiry

International Nuclear Information System (INIS)

Barnes, M.

1983-01-01

Of the four arguments used in favour of Sizewell, the ECC considers that the economic benefit case is much the most important. Consumers will need to be convinced above all that Sizewell will help to restrain the growth of electricity prices. Of the other three arguments, plant need is not taken very seriously by the Board, making provision for future investment options hardly appears in the Statement of Case, and fuel diversity - while an important objective - can only be examined properly within the context of long-term investment strategy and relates only marginally to Sizewell as a self-standing project. Because a decision about Sizewell can in any case only be fully evaluated within the context of long-term plant strategy, the ECC has also briefly considered the most important long-term issues - the existence or otherwise of a nuclear programme, the questions surrounding technology acquisition and alternative energy investments - against which the Sizewell decision should be viewed. (author)

Sizewell 'B' power station public inquiry. CEGB proof of evidence

International Nuclear Information System (INIS)

Flint, R.A.

1982-11-01

The NNC cost estimates for the construction of the Sizewell B nuclear power station are discussed and the methods employed in their preparation are described. The preparation of the estimates for each of the major divisions of cost are outlined. These divisions are: civil engineering and building works; nuclear steam supply system; NNC costs; turbine generators; other mechanical plant; other electrical plant. The uncertainties which exist in making these estimates are discussed and the cost allowances which must be made as a result are outlined. (U.K.)

A RETRAN-02 model of the Sizewell B PCSR design - the Winfrith one-loop model, version 3.0

International Nuclear Information System (INIS)

Kinnersly, S.R.

1983-11-01

A one-loop RETRAN-02 model of the Sizewell B Pre Construction Safety Report (PCSR) design, set up at Winfrith, is described and documented. The model is suitable for symmetrical pressurised transients. Comparison with data from the Sizewell B PCSR shows that the model is a good representation of that design. Known errors, limitations and deficiencies are described. The mode of storage and maintenance at Winfrith using PROMUS (Program Maintenance and Update System) is noted. It is recommended that users modify the standard data by adding replacement cards to the end so as to aid in identification, use and maintenance of local versions. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Mawer, W.T.

1982-11-01

The cooling water system which would be needed for the proposed Sizewell B power station is described. The possible effects upon public safety, local fishing interests and the existing seabed are examined. Studies of the heat dispersion from outfalls, the location of offshore structures and the control of marine growth in the system are reported. Finally the essential service water system, the reserve ultimate heat sink, the discharge of liquid radioactive waste into the sea and the possible use of waste heat for agricultural and fish farming purposes are discussed. (U.K.)

Sizewell B - analysis of British application of US PWR technology

International Nuclear Information System (INIS)

1983-05-01

This report provides information on the staff's evaluation of major design differences and issues developed by the British in their application (Sizewell B) of US PWR technology. One design change, the addition of steam-driven charging pumps, was assessed to have a relatively high value compared to the other changes. However, the assessment is based on a number of assumptions for which inadequate data exist to make an unqualified judgment. Other changes to the US design (as typified by the SNUPPS design) were found to have relatively low or moderate safety benefits for US application

Inpile honing of Sizewell primary selector valve housings

International Nuclear Information System (INIS)

Grindrod, A.; Ward, R.G.

1976-03-01

Difficulties have been experienced at Sizewell power station with the removal and replacement of several of the primary selector valves fitted in the reactors, during the annual maintenance programme. An inpile honing device is described which was specifically designed and developed to facilitate the restoration of the inner sealing faces of the valve housings. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Arnold, P.J.

1982-11-01

The systems benefits of the selection of the Sizewell site for a PWR power station are dealt with. The transmission modifications which would be needed to provide effective connection of this station to the system are considered. (U.K.)

Pressurised water reactor operation

International Nuclear Information System (INIS)

Birnie, S.; Lamonby, J.K.

1987-01-01

The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

Sizewell: good value for consumers' money

International Nuclear Information System (INIS)

MacKerron, G.

1984-01-01

For consumers, the overriding issues in electricity policy are always likely to be those which affect prices, both in the near future and over the longer term. It is therefore the 'need and economics' part of the overall case for Sizewell that consumers' representatives chose to scrutinize. This paper argues that there are no compelling short-term reasons to start construction of Sizewell at the earliest possible time. Further, long-term considerations suggest that there are good reasons not to make an early commitment to the Westinghouse four-loop design of PWR, either for Sizewell or for subsequent orders. (author)

Operational experience with the Sizewell B integrated plant computer system

International Nuclear Information System (INIS)

Ladner, J.E.J.; Alexander, N.C.; Fitzpatrick, J.A.

1997-01-01

The Westinghouse Integrated System for Centralised Operation (WISCO) is the primary plant control system at the Sizewell B Power Station. It comprises three subsystems; the High Integrity Control System (HICS), the Process Control System (PCS) and the Distributed Computer system (DCS). The HICS performs the control and data acquisition of nuclear safety significant plant systems. The PCS uses redundant data processing unit pairs. The workstations and servers of the DCS communicate with each other over a standard ethernet. The maintenance requirements for every plant system are covered by a Maintenance Strategy Report. The breakdown of these reports is listed. The WISCO system has performed exceptionally well. Due to the diagnostic information presented by the HICS, problems could normally be resolved within 24 hours. There have been some 200 outstanding modifications to the system. The procedure of modification is briefly described. (A.K.)

Issues regarding transient analysis examined by the Sizewell B Public Inquiry

International Nuclear Information System (INIS)

Farmer, P.R.; Dunnicliffe, C.J.

1988-01-01

Issues on PWR safety transient analysis that were discussed at the Sizewell B Public Inquiry are presented. The Public Inquiry was set up by the UK Government under an Inspector, Sir Frank Layfield, to examine all aspects of the construction, safety and operation of a 1200 MW(e) PWR on the Sizewell site. The terms of reference were broad ranging, and the constitution of the Inquiry was to make a recommendation under three Acts of Parliament which apply to the construction and operation of nuclear electrical plant. The Inquiry also covered local planning aspects, which are the responsibility of the Local Authority - in this case the Suffolk County Council. The Inspector examined and made recommendations on the safety of the Station, but consideration by Public Inquiry is outside the formal safety and licensing process, which is the business of the Utility (the CEGB) and the Nuclear Installations Inspectorate (the NII). The paper therefore takes a broader look at the question of safety, dealing with the licensing process, the requirements of the safety case and the forward strategies adopted by the CEGB in terms of research and development. This is considered for transient analysis, and the aim is to set the discussions and conclusions of the Public Inquiry into their proper context with regard to nuclear safety in the UK. The Inquiry went into some depth on the topic of LOCA, as an example of safety analysis. In the summary of the evidence and cross-examination the Inspector accepted the adequacy of the LOCA safety case without major reservations, and was satisfied further work in progress would resolve any residual criticisms. In particular support was given for the CEGB commitment to the development and use of more physically realistic calculational methods

Sizewell nuclear power station: investigation of radiation exposure pathways from liquid effluents. Local habits survey 1981

International Nuclear Information System (INIS)

Leonard, D.R.P.; Smith, B.D.

1982-01-01

A habits and consumption survey to review radiation exposure pathways due to liquid effluents released from the CEGB Sizewell site is described. It is relevant to both the Sizewell A and proposed Sizewell B nuclear power stations. The main objectives are to provide input data to a radiological assessment by means of identifying critical groups and to provide data for guidance in a review of environmental monitoring programmes. The way in which data for the different pathways should be combined in order to aid the subsequent radiological assessment is discussed. Recommendations are made for adjustments to the present monitoring programmes. (U.K.)

Showcase or swansong: the dilemma of Sizewell B

International Nuclear Information System (INIS)

Ashmore, Colin.

1996-01-01

The proposal to privatize the United Kingdom's nuclear power generators (Nuclear Electric and Scottish Nuclear, now joined and renamed British Energy) is driven by political rather than rational motives, it is argued. Notwithstanding that nearly one third of the current United Kingdom generating capacity, comes from British Energy, the future looks uncertain, and plans for future nuclear plants at Hinkley Point and Sizewell have been withdrawn, rendering the long-term future of the nuclear fission based industry very unattractive to investors. As gas prices rise, in the long term, the economic balance is likely to change again, with nuclear fusion technology, most likely to benefit. (UK)

Sizewell B

International Nuclear Information System (INIS)

1983-07-01

The concern for failure of the reactor pressure vessel arises not only from the fact that rupture of the vessel may produce a loss of coolant accident in excess of the design basis for the emergency core cooling system, but may simultaneously give rise to missiles which could damage other protection or safeguards systems, including the containment. The safety case for the reactor pressure vessel is discussed with reference to design, manufacture, fracture analysis, inspection and operation. Outstanding issues are identified. (UK)

Load shedding and emergency load sequencing system at Sizewell B power station

International Nuclear Information System (INIS)

Bowcock, S.; Miller, D.

1992-01-01

Sizewell B Nuclear Power Station has a main electrical system that connects together the main turbo-generators, generating at 23.5kV, the 400kV grid and the auxiliary equipment required to operate the station. A separate essential electrical system fed from the main electrical system, supplies all the auxiliaries required to shut-down the nuclear reactor and maintain it in a safe shut-down condition. For safety reasons four similar independent essential electrical systems are provided, each headed by a 3.3kV switchboard and a stand-by 8MW diesel generator. Feeds from the 3.3kV switchboards in turn supply the essential 3.3kV drives and transformer fed 415V essential switchboards. The function of the Load Shedding and Emergency Load Sequencing (LSELS) System is to monitor the condition of the 3.3kV incoming supply from the main electrical system to each essential 3.3kV switchboard and initiate its replacement, with the supply from the associated diesel generator, if it is outside set parameters. In order to achieve this transfer the essential electrical system load must be reduced to a level which the diesel can accommodate as a standing load and then allow the sequenced reconnection of required loads so as not to overload the diesel. The LSELS equipment is categorised as Safety Category 1E and has a significant importance to the safe operation of the power station. Therefore the design of the system must be highly reliable and the purpose of this paper is to detail the design approach used to ensure that a high system reliability is achieved. (Author)

How has severe accident analysis contributed to sizewell B and how can it continue to contribute in the future

International Nuclear Information System (INIS)

Harrison, J.R.; Western, D.J.

1987-01-01

Sizewell B is a proposed 1100 MWe PWR which is a UK development of the US SNUPPS design. The UK reference design document for the plant was first issued in 1981 and the Pre-Construction Safety Report (PCSR) was submitted to the Nuclear Installations Inspectorate (NII), the UK licensing authority, in 1982. A major public inquiry into the proposal took place between January 1983 and March 1985. This paper is concerned with the analysis of severe accidents. This means all the analysis that is concerned with those fault sequences that are outside the design basis of the plant and which may lead to severe consequences - either in terms of plant damage or release of radioactivity. This analysis comprises probabilistic assessments of the frequency of such sequences, transient analysis of the way such sequences develop and radiological release analysis. Part one of this paper examines how the severe accident analysis carried out for Sizewell B has contributed to the judgement that the design is sound and that the construction phase should proceed. The second part of the paper looks to the future and asks ''Can severe accident analysis make any further contribution during the period from licensing up until operation commences

United Kingdom and USSR reactor types

International Nuclear Information System (INIS)

Lewins, Jeffery

1988-01-01

The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

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

International Nuclear Information System (INIS)

1980-10-01

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

The construction of a PWR power station reactor building liner

International Nuclear Information System (INIS)

Skirving, N.; Goulding, J.S.; Gibson, J.A.

1991-01-01

Cleveland Bridge and Engineering Co Ltd (CBE) are constructing the Reactor Building Liner Plate containment of the Sizewell 'B' Power Station for Nuclear Electric Ltd. This has entailed extensive offsite prefabrication of components and their subsequent erection at Sizewell. It has been necessary to engineer temporary supporting mechanisms to enable manufacture and erection to proceed, yet also to withstand wet concrete forces during the progressive construction. The Reactor Building Liner Plate is a safety related system and as such, in addition to strict compliance with the ASME code, the Quality Assurance (QA) requirements of BS 5882 are applicable. A dedicated Project Team was established by CBE to control and direct the work. Equally important as satisfying the rigorous Q.A. requirements has been the need to meet programme and budget. This paper details CBE execution of the Project. (author)

Sizewell: UK power demand

International Nuclear Information System (INIS)

Anon.

1986-01-01

The Sizewell Inquiry was about whether the next power stations to be built in the UK should be nuclear or coal and, if nuclear, PWRs or AGRs. During the period of the Inquiry forecasts of demand for electricity were low. Now, however, it seems that the forecast demand is much increased. This uncertainty in demand and the wide regional variations are examined in some detail. Facts and figures on electricity sales (area by area) are presented. Also the minutes of supply lost per consumer per year. These show that security of supply is also a problem. It is also shown that the way electricity is used has changed. Whilst electricity generation has been changing to large-scale, centralised power stations the demand patterns may make smaller scale, quickly-constructed units more sensible. The questions considered at the Sizewell Inquiry may, indeed, no longer be the right ones. (UK)

Sizewell B

International Nuclear Information System (INIS)

1983-07-01

The safety of pressure circuit components is discussed with respect to coolant loop piping, coolant pumps, pressuriser, steam generator channel head and shell, primary component supports and restraints, accumulators, main steam line no-break zone, reactor internals and core, and valves including main steam isolating valves. Outstanding issues are identified and a programme of additional work is discussed. (U.K.)

An examination of the proposals for the off-site electrical power sources at the Sizewell B PWR

Energy Technology Data Exchange (ETDEWEB)

Woodhouse, P. A. [HM Nuclear Installations Inspectorate, London (United Kingdom)

1986-02-15

Over the past few years there has been an increase in the attention being given to the adequacy and reliability of alternative sources of power provided to supply safety equipment should off-site electrical sources fail. This paper discusses the rationale of HM Nuclear Installations Inspectorates assessment of the electrical systems proposed for the UK's first Pressurized water Reactor, Sizewell 3. The requirements for on-site sources are given, and a discussion is provided of the NII's Assessment Principles including common mode failure, single failure criterion and reliability targets. Where the assessment has resulted in notifications to the original design the reasons are given. The UK's large interconnected Grid System makes complete losses of off-site power comparatively rare. The potential exists however and this paper shows how the current approach ensures that not only are adequate on-site sources available but also that their siting, maintenance and testing are such that loss of off-site power will not cause an unacceptable risk to the public. (author)

Ballooning analysis for the Sizewell B PWR using symmetric MABEL calculations

International Nuclear Information System (INIS)

Sweet, D.W.; Gibson, I.H.; Fell, J.

1982-12-01

An analysis of the fuel clad ballooning potential associated with the Sizewell B PWR following a design basis large break cold leg LOCA is described. Calculations employ MABEL-2C code. No allowance has been made for asymmetries in power or geometry, thus precluding any amelioration offered by early clad rupture. Thermal hydraulic data were derived from a TRAC-PD2 best estimate analysis of the LOCA and the work includes a detailed sensitivity study which leads to a correlation between peak clad temperature and clad strain. For the best estimate start of cycle 1 peak rod rating, no loss of coolability is expected within 95 percent confidence limits on peak clad temperature. No loss of coolability is expected either for rods at the design basis peak rod rating. The temperature does not have to be much higher than the 95 percent confidence limit on the best estimate rating or much beyond that of the design basis rating for rod contact and severe blockage to follow. This indicates that to establish a complete safety case the added complexity of pellet eccentricity and rod to rod power variations must be considered. (U.K.)

Sizewell B cycle 5 core design with Framatome ANP's CASCADE-3D and British Energy's PANTHER

International Nuclear Information System (INIS)

Attale, F.; Koegl, J.; Knight, M.; Bryce, P.

2001-01-01

Sizewell B Cycle 5 is the first cycle, after 4 cycles with BNFL fuel, with a reload consisting of Framatome ANP HTP (high thermal performance) fuel assemblies. The impact of this fuel vendor change on the Nuclear Design area is that, according to British energy's (BE) practice, the Framatome ANP's nuclear design code system CASCADE-3D is used for the majority of the cycle specific safety case calculations. However, other parts of the safety submission (e.g. 3D transient analyses) are made by using the BE code PANTHER. Before using in parallel two different code systems for reload core licensing extensive comparisons of applied methodologies and obtained results were required to ensure an acceptable level of agreement. (orig.)

Lessons from the Sizewell B inquiry: how to make major public inquiries into energy projects fair and efficient. Address to the parliamentary group for energy studies

Energy Technology Data Exchange (ETDEWEB)

O' Riordan, T.; Purdue, M.; Kemp, R.

1986-07-01

The paper is an address to the Parliamentary Group for Energy Studies, and concerns an appraisal of the Sizewell B Inquiry. The unique nature of the Inquiry is described, and an assessment of the Inquiry is given. Based on the main criticisms of the Inquiry, proposals for the reform of future major public inquiries are put forward.

Inspection of the Sizewll 'B' reactor coolant pump flywheels

International Nuclear Information System (INIS)

McNulty, A.L.; Cheshire, A.

1992-01-01

The Sizewell ''B'' safety case has categorised some primary circuit items as components for which failure is considered to be incredible. These Incredibility of Failure (IOF) components are particularly critical in their safety function, and specially stringent and all embracing provisions are made in their design, manufacture, inspection and operation. These provisions are such as to limit the probability of failure to levels which are so low that it does not have to be taken into account and no steps are necessary to control the consequences. The reactor coolant pump flywheel is considered to be an IOF component. Consequently there is a need for rigorous inspection during both manufacture and in service (ISI). The ISI requirement results in the need for an automated inspection. There is therefore a prerequisite to perform a Pre-Service Inspection (PSI) for baseline fingerprinting purposes. Furthermore there is a requirement that the inspection procedure, the inspection equipment and the operators are validated at the Inspection Validation Centre (IVC) of the AEA Technology laboratories at Risley. Development work is described. (author)

Agricultural production data for the Sizewell area

International Nuclear Information System (INIS)

1982-11-01

The proposed site for the Sizewell B nuclear power station is located at the centre of the East Suffolk coastline. The county itself is predominantly agricultural, producing a very wide range of foodstuffs which in the case of one or two particular foodstuffs contribute a significant proportion of the total national production. A general view of the agricultural setting within which the development would be sited and a compendium of more detailed agricultural data relating to the immediate vicinity of the site, which has been drawn on in the Ministry's assessment of the radiological impact of routine atmospheric waste emission, are given. (U.K.)

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

International Nuclear Information System (INIS)

Woollam, P.B.

1984-08-01

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

Democracy and the Sizewell inquiry

International Nuclear Information System (INIS)

Armstrong, J.

1987-01-01

In this paper the brief summary of selected procedural aspects of the Sizewell Inquiry indicates that the way in which major and controversial planning proposals are subject to public debate is in need of urgent review. The Sizewell Inquiry fell short of public expectation, and indeed of Government promises, on two major counts; it did not provide a forum for the debate of certain questions which are clearly of utmost importance and concern to the nuclear issue, while its semi-judicial format and unbalanced funding effectively discouraged and often prevented the level of participation which many objecting parties sought. The Government's assurance of a 'full and fair' debate proved to be hollow and in some quarters this has served to increase scepticism of the stated desire for public involvement in nuclear decision-making. Of the major controversies which have been in the subject of public inquiries in recent years, nuclear power is clearly one of the most challenging. It brings together an extremely varied band of objectors which, in the light of recent radioactive leakages from Windscale, future plans for waste storage facilities and the proposed plutonium reprocessing plant at Dounreay, promises only to grow

Sizewell B cycle 5 core design with Framatome ANP's CASCADE-3D and British Energy's PANTHER

Energy Technology Data Exchange (ETDEWEB)

Attale, F.; Koegl, J. [Framatome ANP GmbH, Nuclear Fuel Cycle, Erlangen (Germany); Knight, M.; Bryce, P. [British Energy, Nuclear Technology Branch, Gloucester (United Kingdom)

2001-07-01

Sizewell B Cycle 5 is the first cycle, after 4 cycles with BNFL fuel, with a reload consisting of Framatome ANP HTP (high thermal performance) fuel assemblies. The impact of this fuel vendor change on the Nuclear Design area is that, according to British energy's (BE) practice, the Framatome ANP's nuclear design code system CASCADE-3D is used for the majority of the cycle specific safety case calculations. However, other parts of the safety submission (e.g. 3D transient analyses) are made by using the BE code PANTHER. Before using in parallel two different code systems for reload core licensing extensive comparisons of applied methodologies and obtained results were required to ensure an acceptable level of agreement. (orig.)

Automated ultrasonic shop inspection of reactor pressure vessel forgings

International Nuclear Information System (INIS)

Farley, J.M.; Dikstra, B.J.; Hanstock, D.J.; Pople, C.H.

1986-01-01

Automated ultrasonic shop inspection utilizing a computer-controlled system is being applied to each of the forgings for the reactor pressure vessel of the proposed Sizewell B PWR power station. Procedures which utilize a combination of high sensitivity shear wave pulse echo, 0 degrees and 70 degrees angled longitudinal waves, tandem and through-thickness arrays have been developed to provide comprehensive coverage and an overall reliability of inspection comparable to the best achieved in UKAEA defect detection trials and in PISC II. This paper describes the ultrasonic techniques, the automated system (its design, commissioning and testing), validation and the progress of the inspections

Sizewell integrated system for centralized operations: getting it all together

International Nuclear Information System (INIS)

Cook, B.M.; Robinson, D.E.P.

1992-01-01

In June 1991, Westinghouse was awarded a contract by Nuclear Electric to supply the Integrated System for Centralized Operations (ISCO) for the Sizewell B nuclear power plant under construction in Suffolk. This system is a large scale distributed control and instrumentation system that will perform safety and control functions for the primary and secondary systems for the plant. Combined with the control and instrumentation equipment that was already being supplied for the plant by Westinghouse, this will be the most extensive integration of modern digital control equipment on a nuclear power plant to date. The ISCO architecture, consisting of 3 major segments -high integrity control, process control and the distributed computer system - is described. The history of the ISCO development is explained. (Author)

Comparison of pre-test analyses with the Sizewell-B 1:10 scale prestressed concrete containment test

International Nuclear Information System (INIS)

Dameron, R.A.; Rashid, Y.R.; Parks, M.B.

1991-01-01

This paper describes pretest analyses of a one-tenth scale model of the 'Sizewell-B' prestressed concrete containment building. The work was performed by ANATECH Research Corp. under contract with Sandia National Laboratories (SNL). Hydraulic testing of the model was conducted in the United Kingdom by the Central Electricity Generating Board (CEGB). In order to further their understanding of containment behavior, the USNRC, through an agreement with the United Kingdom Atomic Energy Authority (UKAEA), also participated in the test program with SNL serving as their technical agent. The analyses that were conducted included two global axisymmetric models with 'bonded' and 'unbonded' analytical treatment of meridional tendons, a 3D quarter model of the structure, an axisymmetric representation of the equipment hatch region, and local plane stress and r-θ models of a buttress. Results of these analyses are described and compared with the results of the test. A global hoop failure at midheight of the cylinder and a shear/bending type failure at the base of the cylinder wall were both found to have roughly equal probability of occurrence; however, the shear failure mode had higher uncertainty associated with it. Consequently, significant effort was dedicated to improving the modeling capability for concrete shear behavior. This work is also described briefly. (author)

Comparison of pre-test analyses with the Sizewell-B 1:10 scale prestressed concrete containment test

International Nuclear Information System (INIS)

Dameron, R.A.; Rashid, Y.R.; Parks, M.B.

1991-01-01

This paper describes pretest analyses of a one-tenth scale model of the Sizewell-B prestressed concrete containment building. The work was performed by ANATECH Research Corp. under contract with Sandia National Laboratories (SNL). Hydraulic testing of the model was conducted in the United Kingdom by the Central Electricity Generating Board (CEGB). In order to further their understanding of containment behavior, the USNRC, through an agreement with the United Kingdom Atomic Energy Authority (UKAEA), also participated in the test program with SNL serving as their technical agent. The analyses that were conducted included two global axisymmetric models with ''bonded'' and ''unbonded'' analytical treatment of meridional tendons, a 3D quarter model of the structure, an axisymmetric representation of the equipment hatch region, and local plan stress and r-θ models of a buttress. Results of these analyses are described and compared with the results of the test. A global hoop failure at midheight of the cylinder and a shear/bending type failure at the base of the cylinder wall were both found to have roughly equal probability of occurrence; however, the shear failure mode had higher uncertainty associated with it. Consequently, significant effort was dedicated to improving the modeling capability for concrete shear behavior. This work is also described briefly. 5 refs., 7 figs

The choice of cement for the manufacture of concrete to be activated: the potential for reducing the radiological consequences of reactor decommissioning

International Nuclear Information System (INIS)

Woollam, P.B.

1985-05-01

This report presents trace element analyses of some candidate cements which might be used in the manufacture of Sizewell 'B' concrete. It completes a programme of work whose aim was to investigate the potential for reducing the radiological consequences of reactor decommissioning through selection of construction materials for activated components. In particular, consideration has been given to the potential for reducing the concentration of elements known to activate to long lived daughters. (U.K.)

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

A new reactor core monitoring system. First experience gained at the Dukovany NPP

International Nuclear Information System (INIS)

Pecka, M.; Svarny, J.; Kment, J.

2001-01-01

The article deals with methods of interpretation of in-core measurements that are based on the determination of the three-dimensional (3D) power distribution within the reactor core, discusses on-line mode calculations, and describes the results obtained during the trial operation of the new SCORPIO-VVER reactor core monitoring system. The principles of the method of determination of the fuel assembly subchannel parameters are outlined. Alternative methods of self-powered detector signal conversion to local power are given, and some results of their testing are presented. Emphasis is put on self-powered detectors supplied by the US firm IST, which were first deployed at the Dukovany NPP in 1998. The predictive function of the SCORPIO-VVER system, whose implementation was inspired by favourable experience gained on some PWR reactors (such as the products of the Halden reactor project at Ringhals and Sizewell B) were adapted to the specific needs of WWER-440 reactors. The main results of validation of the functions are described and presented in detail. (author)

The regulatory approach to the assessment of containment structures in the United Kingdom, with reference to the proposed Sizewell 'B' project

International Nuclear Information System (INIS)

Todd, I.W.

1988-01-01

The regulation of nuclear installations in the United Kingdom is carried out by means of a licensing system which is administered on behalf of the Health and Safety Executive (HSE) by HM Nuclear Installations Inspectorate. A major part of this process lies in the assessment of an installation's designs and safety case, which must be submitted to the Inspectorate before a licence may be granted. This paper describes the work carried out in assessing the design proposed for the containment structure of the United Kingdom's first PWR, Sizewell 'B'. The roles of the Inspectorate's Safety Assessment Principles, and the engineering judgement of the individual assessor, are discussed. The reasons behind the Inspectorate's decision to commission independent design studies of the structure are explained, and the results of these studies are presented in outline. Views are also expressed upon the validation of analytical results, quality assurance, and further work planned by the Inspectorate in this field. The paper concludes that it is the flexibility and discretion allowed by the above approach which represents the principal strength of the UK's system of regulatory assessment. (author)

The effects of a severe reactor accident at the proposed Sizewell B station upon agriculture and fisheries in the United Kingdom and neighbouring countries

International Nuclear Information System (INIS)

Taylor, P.J.

1984-05-01

This report is a preliminary assessment of the effects of a degraded core accident at the Sizewell site upon regional, national and international agricultural production and fisheries. Two scales of release are taken, one where the containment works effectively, and one where it is breached within a few hours. These two accidents correspond to the UK-11 and UK-1 releases studied by the NRPB. Hitherto, accident consequence studies have focussed upon numbers of cancers or radiation deaths as a measure of the impact. There have been no detailed studies of the impact of the radioactive fall-out upon agricultural production itself, or upon fisheries if the releases are carried out to sea. This report provides a first assessment of the areas of land contaminated by the two scales of accident under weather conditions representative of the range that occurs normally at the site and for a number of different wind directions. The results show that if the containment remains intact, effects would be confined to the locality, with restrictions extending little further than 30 km. However, in the case of the most severe release, agricultural restrictions would be extensive, depending upon wind direction, over the whole of the British Isles. (author)

Radioactivity in the vicinity of Sizewell nuclear power station: marine environmental monitoring, 1983

International Nuclear Information System (INIS)

Hunt, G.J.

1984-01-01

Regular monitoring of radioactivity in the aquatic environment of the British Isles has continued. A report presents the results of this program in relation to the Sizewell nuclear power station for 1983, so as to supplement the most recently available full report which is for 1982. (author)

Achilles tests finally nail PWR fuel clad ballooning fears

International Nuclear Information System (INIS)

Dore, P.; McMinn, K.

1992-01-01

A conclusive series of experiments carried out by AEA Reactor Services at its Achilles rig in the UK has finally allayed fears that fuel clad ballooning is a major safety problem for Sizewell B, Britain's first Pressurized Water Reactor. The experiments are described in this article. (author)

Hanford B Reactor Building Hazard Assessment Report

International Nuclear Information System (INIS)

Griffin, P. W.

1999-01-01

The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

Friends of the Earth case against the PWR

International Nuclear Information System (INIS)

Boyle, S.

1987-01-01

Friends of the Earth's case against Sizewell B has been summarised in a report entitled 'Critical Decision: Should Britain buy the Pressurised Water Reactor?'. This showed that on economic and safety grounds, Sizewell B would not be a good choice for the electricity consumer or the country at large. Events since the end of the Inquiry, particularly those affecting the economic case, have confirmed this conclusion. This paper will summarise the case, both during the Inquiry and subsequent to this, as well as make reference to the long-term environmental implications of the Central Electricity Generating Board's PWR programme. (author)

The place of nuclear power on the CEGB system

International Nuclear Information System (INIS)

Jenkin, F.P.

1987-01-01

This paper discusses the developments in the Central Electricity Generating Board's programme of nuclear generation of electricity since the Sizewell-B inquiry. In particular, recent trends in electricity demand growth are discussed. An average growth in both electricity sales and peak demand of over 1.5%pa during the next ten years is predicted, so that by the mid-1990s an additional 5GW capacity will be needed. Energy conservation is one factor taken into consideration when estimating demand. The need for new capacity is illustrated graphically. With the closure of old plant the new generating capacity needed by 2000 is estimated at 13GW with Sizewell-B supplying 1.2GW of this. This demand will be met by construction of both nuclear and coal fired stations. As the demand increase is greater in the south of Britain than in the north, more of this new generating capacity will be built in the south. The CEGB policy is to build four or five Pressurized Water Reactors the same as Sizewell-B. The sites under consideration for new nuclear stations are shown. They are Hinkley Point, Winfrith, Dungeness, Sizewell, Trawsfynnedd, Wylfa and Druridge. (UK)

The 10B(n,α)7Li reaction in PWR coolants: calculations of the effect on coolant pH and on decreases in 10B isotopic fractions

International Nuclear Information System (INIS)

Polley, M.V.

1988-07-01

Boron is used as a chemical shim in PWRs for reactivity control and is added in the form of boric acid to the primary coolant. The 10 B(n,α) 7 Li reaction leads to a continuous increase in 7 Li in the primary coolant and to a continuous decrease in 10 B the isotope of boron responsible for control of reactivity. The rate of increase in coolant pH due to 7 Li production is calculated for the Sizewell 'B' PWR to enable judgements to be made on the frequency of sampling and removal of lithium required to maintain the pH of the primary coolant within the desired limits. Calculations are contrasted for the cases of natural boron and 100% 10 B chemical shims, for both a normal cycle and an extended 18 month cycle. Calculations of 10 B depletion over 30 years of operation as a function of the quantity of boron discharged to waste are also presented. 10 B isotopic fractions are calculated for the reactor coolant (RC), boric acid tanks (BATs) and refuelling water storage tank (RWST) assuming rapid mixing of BAT and RC boron for tritium control and other reasons. Such predictions enable assessments of the reactor physics implications of 10 B consumption to be made. (author)

Risk assessment for LRW

International Nuclear Information System (INIS)

Anon.

1987-01-01

The information obtained by probabilistic methods is explained discussing the results of various studies, such as the Reactor Safety Study of the USA, the West-German Reactor Safety Study (phases A and B), some other important studies like the RSSMAP and the IREP, as well as the Sizewell B report and an NRC report on precursors to potential severe core damage accidents. (DG) [de

Frequency limitations on reactor release derived from radiological criteria

International Nuclear Information System (INIS)

Clarke, R.H.

1976-05-01

Acceptable frequencies are suggested with which a number of defined individual and collective dose criteria are met. This has enabled a relationship to be proposed between the frequency and any event leading to release of radioactive material and the dose commitment to the hypothetical individual at the site boundary. The philosphy developed is generally applicable, but in this instance it has been demonstrated for an SGHW Reactor located at Sizewell. The results are derived on the basis of an acceptable level of risk of fatality to the most exposed individual and the application of the methodology to other reactor types and sites is demonstrated. (author)

Hinkley Point 'C' power station public inquiry: proof of evidence on safety criteria

International Nuclear Information System (INIS)

Taylor, R.H.

1988-09-01

A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. The policy is to replicate the Sizewell ''B'' PWR design which was accepted as safe by an earlier enquiry. In this evidence to the Inquiry, subsequent developments are examined with a view to determining whether these would reverse the Sizewell decision. They are: the possible revision of radiation risk estimates upwards; whether cases of leukaemia occur with greater frequency around nuclear sites than elsewhere; publication of the Health and Safety Executive's consultative document ''The Tolerability of Risk from Nuclear Power Stations''. The overall conclusion is that these developments do not undermine the findings of the Sizewell ''B'' inquiry or the validity of the CEGB's safety criteria. (author)

The effect of fish impingement at Sizewell 'A' Power Station, Suffolk, on North Sea fish stocks

International Nuclear Information System (INIS)

Turnpenny, A.W.H.; Utting, N.J.; Millner, R.S.; Riley, J.D.

1988-04-01

Samples collected from the cooling water intake screens of Sizewell 'A' power station over a 12 month period contained 73 species of fish. Of these, only 20 species were present on more than 50% of sampling dates and only 7 commercially exploited species were caught in quantities of more than a few hundred over the year; namely sprat, herring, cod, whiting, sole, dab and plaice. These species formed the basis of analysis of the impact of the Power Station on commercial species. Commercial species found in the Sizewell area are part of major North Sea stocks. The impact of the losses due to the Power Station is spread over these stocks, hence the effect is minimal. The mortality rate caused by the Power Station is one thousandth to one hundred-thousandth, depending on species, of that caused by commercial fishing and the effect is less than that of a small, inefficient commercial trawler. (author)

PWR design for low doses in the United Kingdom: The present and the future

Energy Technology Data Exchange (ETDEWEB)

Zodiates, A.M.; Willcock, A. [PWR Project Group, Knutsford, England (United Kingdom)

1995-03-01

The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B, presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.

Hinkley Point 'C' power station public inquiry: proof of evidence on design and safety

International Nuclear Information System (INIS)

George, B.V.

1988-09-01

A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. The policy is to replicate the Sizewell ''B'' PWR design. The Hinkley Point ''C'' design is described indicating where changes in the Sizewell ''B'' design have been made to accommodate site differences. These are associated with the civil engineering construction and some of the electrical systems and do not affect the safety case. External hazards differ from site to site and the effect on the safety case of those specific to Hinkley Point are examined. The Chernobyl accident and the assessment of the United Kingdom PWR which was carried out subsequently are reviewed. The assessment indicated that no changes in the Sizewell ''B'' design and safety case were called for as a result of this accident; accident management developments are also reviewed, however. The CEGB's approach to minimizing occupational radiation doses is described. (UK)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Wright, J.K.

1982-11-01

A description is given of the CEGB's current assessment of the potential contribution to electricity generation of those methods considered to be the most significant alternatives to conventional power stations or thermal nuclear reactors for generating electricity. The state of development of the technology, the timescale on which it could be developed, its ability to generate electricity reliably and the cost of generating electricity are some of the matters considered. The following topics are covered: potential developments in coal fired generating plant; combined heat and power generation; energy storage; electricity generation from renewable energy sources; fast reactors and fusion. (U.K.)

Nuclear and non-nuclear risks: Address to the Parliamentary and Scientific Committee

International Nuclear Information System (INIS)

Gittus, J.

1988-01-01

Risk assessment is first explained. For example, for a loss of coolant accident at Sizewell B the estimate of the likelihood that radioactivity would be released involves three numbers:- the frequency of loss of coolant from the reactor; how often the safety pumps which would have to work in case of a loss of coolant have failed when tested; and the chance that the containment round the reactor will fail. As each risk number is small, the combined risk is very small. Using the same method of assessment the risk of an accident affecting public safety at the liquid gas plant on Canvey Island is 1000 larger, but still small and acceptable. Much time and money was spent on establishing the 1000 times lower risk of an accident affecting public safety at the Sizewell reactor. The risk assessment was however, finally accepted. Fast reactors are even safer than pressurised water reactors and use a renewable energy source; their development is strongly advocated. It is noted that the risks from a medieval water-wheel were enormous compared with nuclear risks today. (U.K.)

Hinkley Point 'C' power station public inquiry: proof of evidence on potential off-site effects of radiation

International Nuclear Information System (INIS)

Western, D.J.

1988-09-01

A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. This evidence to the Inquiry is concerned with the potential of the proposed Hinkely Point ''C'' PWR to increase the exposure of members of the public offsite to radiation. The policy is to replicate the design of the Sizewell ''B'' reactor. The evidence examined in great detail at the Sizewell ''B'' Public Inquiry where the Inspector concluded that the risk would be very small. The purpose of this evidence is to provide an explicit account of the potential off-site effects of radiation at the Hinkley Point site, so that it can be seen that there is nothing specific to this location that could lead to a different conclusion. (author)

The Sizewell inquiry - is there a better way

International Nuclear Information System (INIS)

Greenhalgh, G.

1984-01-01

The author sees the Inquiry as the latest in a line arising from government's wish to achieve greater public participation in controversial decisions. He believes that the Sizewell Inquiry is unlikely to shake the public out of its apathy, while a decision that goes against them will not satisfy the objectors. The concept of the Inquiry is based on the belief that the legal process will unearth objective truth, while most of the issues are matters of opinion and judgement. The wide-ranging terms of reference are leading to constitutional anomalies and attempts to take on Herculean tasks in the search for objective truth. However, while concluding that an Inquiry Commission adopting a legal approach and following courtroom procedures is not a satisfactory way of dealing with large-scale technological projects, the author finds it hard to suggest alternatives short of more direct parliamentary control. (author)

The power of British Energy

International Nuclear Information System (INIS)

Hawley, R.

1997-01-01

When the power industry in Britain was privatized, British Energy plc (BE), whose head office is in Edingburgh, Scotland, was founded in July 1996. It is the only power utility in the world exclusively operating nuclear power stations. Operative business has remained the responsibility of the two regional supply companies, Nuclear Electric (NE) and Scottish Nuclear (SN) which, in addition to the modern PWR nuclear generating unit of Sizewell B, have included in the new holding company their advanced gas-cooled and gas-moderated reactor (AGR) units. The older gas-graphite reactor (GGR) plants were combined in the new Magnox Electric plc, Berkeley; at some later date, this company is to be merged with another nuclear power plant operator, British Nuclear Fuels plc (BNFL). Sizewell B, which was commissioned in 1995, is the last nuclear generating unit to be started up in the United Kingdom, for the time being. In times of low raw material prices and the need for a quick return on invested capital, BE is reluctant to run the risk associated with tying up capital for a long time. Instead, the company has backfitted its plants so that the production of electricity from nuclear power in Britain in 1996 of 92,476 GWh was increased by almost 10% over the 1995 level of 84,174 GWh. In addition to modernization and rationalization at home, BE together with Sizewell B vendor Westinghouse is engaged worldwide in the development and commercialization of future advanced reactors. This ensures that the know-how accumulated will be preserved and will be available for new nuclear power plants to be built in Britain in the next century. (orig.)

National Nuclear Corporation Limited report and financial statements 31 March 1987

International Nuclear Information System (INIS)

1987-01-01

The paper presents the annual report and audited financial statements of the National Nuclear Corporation Limited, 31 March 1987. The company is engaged in designing and constructing power stations and nuclear power reactors and other related work. A review of the development work carried out on power stations is briefly given, along with the research and development work on reactors. Future prospects concerning the Sizewell-'B' project are briefly outlined. (U.K.)

NNC - the past, present and future

International Nuclear Information System (INIS)

Anon.

1985-01-01

An account is given of the activities of the National Nuclear Corporation and its predecessors. Its historical involvement in the design and construction of all the nuclear power stations currently operated by Britain's generating boards, two stations overseas, and in association with UKAEA the design and construction of the Prototype Fast Reactor at Dounreay, is described. Heysham 2 and Torness, and the Proposed Sizewell B are discussed. Headings are: origins and structure; qualified workforce; choice of reactors; collaborative agreements (on fast reactors); market survey; more diversity. (U.K.)

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

Plans for use of ENDF/B in reactor research in Indonesia

International Nuclear Information System (INIS)

Santoso, B.; Syaukat, A.; Subki, I.; Ganesan, S.

1989-07-01

Nuclear data are numerical constants of nature which quantify the nuclear behaviour of all elements and isotopes which make up the reactor medium and its environment, and which are needed as input for performing design calculations for safe and reliable operation of nuclear reactors. The nuclear data are available in the form of recommended values in specially formatted computerized files such as the Evaluated Nuclear Data File-B, known as ENDF/B. The development of base technology in the scheme of original reactor design calculations involves the mastering of the art of ENDF/B data processing. This paper briefly discusses the current status of this activity in Jakarta and gives an account of the future plans, with emphasis on the role of ENDF/B in reactor calculations. (author). 15 refs, 9 figs

Ergonomics design and operator training as contributors to human reliability

International Nuclear Information System (INIS)

Jackson, A.R.G.; Madden, V.J.; Umbers, I.G.; Williams, J.C.

1987-01-01

The safe operation of nuclear reactors depends not only on good physical safety engineering but on the human operators as well. The Central Electricity Generating Board's approach to human reliability includes the following aspects: ergonomics design (task analysis and the development of man-machine interfaces), analysis of human reliability, operational feedback, staff training and assessment, maintenance management, research programmes and management. This paper describes how these combine to achieve the highest practicable level of human reliability, not only for the Sizewell-B pressurized water reactor, but also for the Board's gas-cooled reactors. Examples are used to illustrate the topics considered. (UK)

Theoretical study for ICRF sustained LHD type p-11B reactor

International Nuclear Information System (INIS)

Watanabe, Tsuguhiro

2003-04-01

This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p- 11 B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D- 3 He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p- 11 B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D- 3 He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor (γ HH ). It is shown that high average beta plasma confinement, a large confinement factor (γ HH > 3) and the hot ion mode (T i /T e > 1.4) are necessary to achieve the ignition of the D- 3 He helical reactor. Characteristics of ICRF sustained p- 11 B reactor are analyzed in section 4. The nuclear fusion reaction rate is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p- 11 B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

Licensing aspects of structural integrity

International Nuclear Information System (INIS)

Turner, M.J.; Hemsworth, B.; Boydon, F.M.D.; Harrop, L.P.; Waters, R.

1992-01-01

Examples are given of the wide variety of structural integrity assessments of nuclear plant carried out by the United Kingdom Nuclear Installations Inspectorate (NII) and the consequent need for a flexible approach within the framework provided by the Safety Assessment Principles. The paper describes the use of the Special Case Procedure and draws the distinction between the assessment of incredibility of failure of components and components whose failures are considered within the design basis. Assessment examples provided are the Sizewell B reactor pressure vessel, Magnox reactor pressure vessels, the Prototype Fast Reactor core support structure, Advanced Gas-cooled Reactor steam plant, Thermal Oxide Reprocessing Plant (THORP) vessels, and Steam Generating Heavy Water Reactor pressure tubes. (author)

Assessing the risks of nuclear energy

International Nuclear Information System (INIS)

Evans, Nigel

1986-01-01

The question how safe is safe is discussed. The way in which nuclear energy is presented in the context of the risks inherent in daily life is considered. Calculations based on actual reactor accidents (not Chernobyl) are then extrapolated for the proposed Sizewell-B reactor. Comparison is made between the risks of the nuclear industry and coal mining. Risk perception is considered and a risk index is constructed that allows for the fact that nuclear power continues to arouse public fear in spite of a good safety record. (UK)

Procedures to relate the NII safety assessment principles for nuclear reactors to risk

CERN Document Server

Kelly, G N; Hemming, C R

1985-01-01

Within the framework of the Public Inquiry into the proposed pressurised water reactor (PWR) at Sizewell, estimates were made of the levels of individual and societal risk from a PWR designed in a manner which would conform to the safety assessment principles formulated by the Nuclear Installations Inspectorate (NII). The procedures used to derive these levels of risk are described in this report. The opportunity has also been taken to revise the risk estimates made at the time of the Inquiry by taking account of additional data which were not then available, and to provide further quantification of the likely range of uncertainty in the predictions. This re-analysis has led to small changes in the levels of risk previously evaluated, but these are not sufficient to affect the broad conclusions reached before. For a reactor just conforming to the NII safety assessment principles a maximum individual risk of fatal cancer of about 10 sup - sup 6 per year of reactor operation has been estimated; the societal ris...

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

The siting of UK nuclear reactors.

Science.gov (United States)

Grimston, Malcolm; Nuttall, William J; Vaughan, Geoff

2014-06-01

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945-1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965-1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985-2005) there was very little new nuclear development, Sizewell

Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Tsuguhiro (ed.)

2003-04-01

This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

Nuclear dewatering

International Nuclear Information System (INIS)

Anon.

1989-01-01

The Sizewell 'B' Nuclear Power Station, the first in a new family of PWR (Pressurised Water Reactor) stations in the UK, is currently under construction on the Suffolk coast. One of the first civil engineering tasks associated with the power station construction was the necessity to lower the water table by 17-18 m over an area of approximately 8 hectares to allow the excavation for the foundations to be carried out in the dry. The way chosen to do this was to construct a diaphragm wall around the site area. This had the least effect on surrounding sites -the Sizewell-A station and various nature reserves and wetlands. The reasons for the choice of method are discussed. Following the construction of the wall the water was pumped out from within the diaphragm wall in January 1988. (author)

Fast-reactor-data testing of ENDF/B-V at ORNL

International Nuclear Information System (INIS)

Wright, R.Q.; Ford, W.E. III; Lucius, J.L.; Webster, C.C.; Marable, J.H.

1982-01-01

The Cross Section Evaluation Working Group (CSEWG) is coordinating a program to assess the adequacy of ENDF/B-V cross sections for both fast- and thermal-reactor design applications. A secondary goal is to evaluate cross-section processing codes, cross-section libraries, and radiation-transport codes. Fast reactor data testing (FRDT) goals are accomplished, in part, by comparison of calculated results with documented performance parameters of CSEWG fast reactor benchmarks and with results obtained by other data testers. The purpose of this paper is to describe the results of FRDT at Oak Ridge National Laboratory

Sizewell Power Station

International Nuclear Information System (INIS)

1985-01-01

A detailed emergency plan for the reactor is presented. Definitions, conditions for taking action, duties of staff, emergency control centres and equipment, communications, both internal and external, hazard assessment, including iodine inhalation and radioactive deposition procedures collaboration with other bodies, warnings to the public and exercises are among topics described. (author)

The Management Advisory Committee of the Inspection Validation Centre seventh report

International Nuclear Information System (INIS)

1990-07-01

The Management Advisory Committee of the Inspection Validation Centre (IVC/MAC) was set up to review the policy, scope, procedure and operation of the Inspection Validation Centre (IVC), to supervise its operation and to advise and report to the United Kingdom Atomic Energy Authority (UKAEA) appropriately. The IVC was established at the UKAEA Risley Laboratory, to validate the procedures, personnel and equipment proposed by Nuclear Electric for use in the ultrasonic inspection at various stages of the fabrication, erection and operation of the Sizewell 'B' Pressurized Water Reactor (PWR) reactor pressure vessel (RPV) and such other components as are identified by the utility. It is operated by the UKAEA to work as an independent organisation under contract to Nuclear Electric, and results are reported to Nuclear Electric together with the conclusions of the Centre in relation to the validation of individual techniques. At the meetings of the IVC/MAC, the progress on the manufacture of the pressure vessel is also outlined by the PWR Project Director. The vessel has now undergone the final stress relief and post-hydro inspection and is due to be delivered to the Sizewell site before the end of 1990. (author)

Report and accounts 1989/90

International Nuclear Information System (INIS)

1991-01-01

Nuclear Electric was set up to run all the Magnox and AGR nuclear power stations and nuclear-related assets of the Central Electricity Generating Board. It is also to be responsible for the construction and operation of the PWR station at Sizewell-B in Suffolk. The decision not to privatise the nuclear section of the electricity supply industry in England but to retain it in government ownership came quite late on so the handover to Nuclear Electric was set up in 20 weeks, so that the new company could start operating on 31st march 1990. This report sets out the first year's progress and reviews the operation of the Magnox and AGR reactors and the construction of Sizewell-B including a revised cost estimation. Health and Safety is an important objective of the company and its committment to this and to environmental protection is stated. The Directors Report and company accounts presented. The accounts show an operating profit but an overall loss for the financial year 1989/90. (UK)

Sizewell 'B' cable installation

International Nuclear Information System (INIS)

Gemmell, D.R.

1992-01-01

N G Bailey and Co. Ltd., UK were awarded the contract for the procurement, manufacture, works testing, works finishing, supply, delivery, off-loading, storage, installation, site finishing, preservation, setting to work and site testing of the following; the main cable installation throughout the Station including the addition of the Radioactive Waste Building, earthing and lightning protection installation, cable supporting steelwork and carriers and glanding and termination of cables. The cabling installation comprises power distribution, control and instrumentation cabling including all the associated cabling accessories, terminal boxes and similar components. The way that the contract was set-up, awarded and is now being carried out is described. Planning and industrial relations have been key features of the contract. (Author)

Graphics and control of the guide tube assembly reinforcement manipulators at Sizewell 'A'

International Nuclear Information System (INIS)

Burden, C.

1996-01-01

A method was devised to reinforce the lower lug welds of the Guide Tube Assemblies (GTA's) at Sizewell 'A'. A six degree of freedom manipulator was designed to place a clamp around the lugs and tighten it. The manipulator was fitted with the three fixed cameras but required another surveillance manipulator positioned in an adjacent standpipe to provide additional views. The need to prepare two standpipes limited the rate at which reinforcements could be made. Therefore an articulated two arm camera manipulator, which could be used on the existing manipulator mast was designed and built. The two manipulators were driven from separate desks and were controlled by the same supervisory computer linked to online graphics. The camera arm joints were driven on preplanned routes using a single joystick because of the complex moves and tight spaces involved. A large number of GTA sites have now been reinforced including a dropped GTA which had to be raised to carry out clamping. (Author)

Preliminary Hazard Classification for the 105-B Reactor

International Nuclear Information System (INIS)

Kerr, N.R.

1997-08-01

This document summarizes the inventories of radioactive and hazardous materials present within the 105-B Reactor and uses the inventory information to determine the preliminary hazard classification for the surveillance and maintenance activities of the facility. The result of this effort was the preliminary hazard classification for the 105-B Building surveillance and maintenance activities. The preliminary hazard classification was determined to be Nuclear Category 3. Additional hazard and accident analysis will be documented in a separate report to define the hazard controls and final hazard classification

A review of the United Kingdom fast reactor programme

International Nuclear Information System (INIS)

Bramman, J.I.; Hickey, H.B.; Whitlow, W.H.; Frew, J.D.; Gregory, C.V.

1990-01-01

Total energy consumption in the UK in 1989 was 340 million tonnes of coal or coal equivalent, made up as follows: coal 31%, petroleum 35%, natural gas 24%, nuclear electricity 8%, hydroelectricity 1% and imported electricity 1%. About half of the nuclear electricity generated came from 14 Advanced Gas-Cooled Reactors (AGRs) and about half from the 24 older gas-cooled Magnox reactors, one Steam-Generating Heavy-Water Reactor (SGHWR) and one fast reactor (the Prototype Fast Reactor, PFR, at Dounreay). The privatization of the Electricity Supply Industry (ESI) in the UK is proceeding. On 9 November 1989, however, it was announced by the Secretary of State for Energy that the privatization plan would be changed and that the CEGB's nuclear stations were to remain in state ownership, through the formation of an additional company, Nuclear Electric. At the same time, the Secretary of State for Scotland announced the formation of a similar state-owned company, Scottish Nuclear. Nuclear Electric was asked, in the interim, to examine priorities in the whole nuclear field with particular reference to the improvement of the economics and performance of existing reactors, to the development of the Sizewell and alternative reactors and to the development of longer-term options such as the fast reactor and fusion. Nuclear Electric has been asked to formulate its new policy by June 1990. The PFR programme will continue to be funded by the UK government until March 1994. AEA Technology is endeavouring to find alternative funding to maintain the operation of the PFR until at least the year 2000. The House of Commons Select Committee on Energy stated in its report that the fast reactor ''is a matter for the British Government to foster as a long-term option for the generation of electricity in this country'', and recommended that in the interim the Government reassesses its position on this new technology in the light of increasing concern about CO 2 emissions and the long

The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

CERN Document Server

Kelly, G N; Charles, D; Hemming, C R

1983-01-01

This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

Calculation and Analysis of B/T (Burning and/or Transmutation Rate of Minor Actinides and Plutonium Performed by Fast B/T Reactor

Directory of Open Access Journals (Sweden)

Marsodi

2006-01-01

Full Text Available Calculation and analysis of B/T (Burning and/or Transmutation rate of MA (minor actinides and Pu (Plutonium has been performed in fast B/T reactor. The study was based on the assumption that the spectrum shift of neutron flux to higher side of neutron energy had a potential significance for designing the fast B/T reactor and a remarkable effect for increasing the B/T rate of MA and/or Pu. The spectrum shifts of neutron have been performed by change MOX to metallic fuel. Blending fraction of MA and or Pu in B/T fuel and the volume ratio of fuel to coolant in the reactor core were also considered. Here, the performance of fast B/T reactor was evaluated theoretically based on the calculation results of the neutronics and burn-up analysis. In this study, the B/T rate of MA and/or Pu increased by increasing the blending fraction of MA and or Pu and by changing the F/C ratio. According to the results, the total B/T rate, i.e. [B/T rate]MA + [B/T rate]Pu, could be kept nearly constant under the critical condition, if the sum of the MA and Pu inventory in the core is nearly constant. The effect of loading structure was examined for inner or outer loading of concentric geometry and for homogeneous loading. Homogeneous loading of B/T fuel was the good structure for obtaining the higher B/T rate, rather than inner or outer loading

Status of data testing of ENDF/B-V reactor dosimetry file

International Nuclear Information System (INIS)

Magurno, B.A.

1979-01-01

The ENDF/B-V Reactor Dosimetry File was released August 1979, and Phase II data testing started. The results presented here are from Brookhaven National Laboratory only, and are considered preliminary. The tests include calculated spectrum-averaged cross sections using 235 U fission spectrum (Watt), 252 Cf spontaneous fission spectrum (Watt and Maxwellian), and the Coupled Fast Reactor Measurement Facility (CFRMF) spectrum. 6 tables

Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

CERN Document Server

Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

1982-01-01

The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-01-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

A Review of Past Reactor Progress in the UK, April 1975

International Nuclear Information System (INIS)

Smith, R.D.

1975-01-01

The fast reactor project in the UK is continuing at approximately the same scale of effort as last year, with great attention being paid to overcoming the difficulties being met in raising PFR to full power. Organisational arrangements for nuclear power in the UK during the year have been dominated by two interlinked issues, firstly the choice of thermal reactor system for the next stage of the British nuclear power programme, and secondly the consolidation of the British nuclear design and construction industry. On the former, widespread discussion on the merits of alternative thermal systems culminated in the announcement by the Secretary of State for Energy on 10 July 1974 that the British electricity boards would adopt the SGHWR for their next nuclear power station orders. The initial ordering programme was set at 4000 KW(E) over the next four years, with the prospect of extending the programme in the later 1970s given satisfactory initial experience of construction. Consent has now been given by the Department of Energy to the CEGB for an SGHWR station at Sizewell in Suffolk (adjacent to an existing Magnox station) and to the SSEB for one at Torness Point in East Lothian

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-11-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

PSA - A utility perspective from the United Kingdom

International Nuclear Information System (INIS)

Ross, P.J.

2004-01-01

This paper provides an outline of the three stages of PSA used for the Sizewell B PWR. These stages cover the use of PSA during design, licensing and (in future) operation of Sizewell B. The paper discusses each of these stages in some detail, highlighting the differences in approach and lessons learnt at each stage. (author)

For sale: 7 AGR stations and a brand new PWR

International Nuclear Information System (INIS)

Anon.

1995-01-01

Britain's seven AGR stations and the Sizewell B PWR will pass to private ownership under the UK government's plan to privatise the two nuclear generators, Nuclear Electric and Scottish Nuclear, sometime next year. Under the new set-up, the two generators will become operating subsidiaries of a holding company which will be headquartered in Scotland. The companies' ageing Magnox gas-cooled reactors will remain in a separate public sector company before being transferred to British Nuclear Fuels (BNFL) at the time of privatisation. (author)

Nuclear power: consolidation or change

International Nuclear Information System (INIS)

1986-02-01

Currently about 45% of the electricity consumed in Scotland comes from nuclear sources and when the power station at Torness, near Edinburgh, is fully commissioned it will be about 60%. Torness is an AGR type reactor and the Scottish Electricty Board (SSEB) consider that there is no case for changing to a PWR type design for any future reactor built in the UK. The economics of 'a next AGR' and a PWR reactor are compared. The construction times for the Torness programme, the electrical output, station life, load factor, fuel costs and additional support costs are also considered, and compared with costs estimated by the Central Elecricity Generating Board (CEGB) for the proposed Sizewell-B reactor. Exposure of station staff to radiation, safety, public acceptibility and development potential, impact on industry, job creation and export potential are also discussed briefly. (UK)

1994 - Starting a new era for nuclear power in Great Britain

International Nuclear Information System (INIS)

Collier, J.G.

2004-01-01

A gap is opening up between what the world would like to consume in terms of energy and what the world's environment can sustain. To bridge this gap a much greater contribution from nuclear and renewables is needed to limit the emission of CO 2 and other harmful gases. Plans already laid in some Pacific Rim countries recognise this need and nuclear energy is expected to grow rapidly in this region over the next 20-30 years. For Pacific Rim countries without plentiful indigenous fossil fuel resources, nuclear power is the most economic form of electricity generation, minimising the need for expensive imports and strengthening security of energy supply. Nuclear power already makes a significant contribution to reducing emissions in other regions, especially Europe. Well over a third of electricity in Europe comes from nuclear power. In the UK the figure is nearer a quarter and rising. If nuclear power is to fulfil its role as a key part of UK energy supply into the 21st century, we need to maintain our existing capacity as our older plant retires by building further, modern PWRs. Central to our future is Sizewell B power station - the UK's first PWR. Construction is complete and commissioning is well under way. The project is nearing completion - ahead of time and under budget. The success of Sizewell B is the foundation for the future of nuclear power in the UK. Stations based on the Sizewell design will become the workhorses of our utility. Already Nuclear Electric has submitted a planning application to construct Sizewell C. This is a technical replica of Sizewell B in twin reactor form with a net output of around 2600MW. And this design also has strong export potential. Westinghouse and Nuclear Electric have submitted a joint-vender to build a PWR here in Taiwan. Our bid is based on the Sizewell design; an innovative, safe and proven nuclear power plant design for supplying reliable low-cost electricity for the Republic of China's growing economy. The Sizewell

Chernobyl

International Nuclear Information System (INIS)

1986-01-01

This leaflet has been prepared by the Central Electricity Generating Board. Following the accident at Chernobyl nuclear power station in the Soviet Union people are concerned about the safety of the UK's nuclear power stations. This leaflet explains that Chernobyl is unlike any nuclear station operating or planned in the UK and under the CEGB's stringent safety rules it could not have been built in the UK. The leaflet explains what happened at Chernobyl and compares the RBMK design and British reactors. The bodies concerned with reactor safety are noted. The containment of radioactivity and emergency procedures are explained. The PWR design for Sizewell-B is stated to be much safer than the RBMK Chernobyl design. (UK)

Reviewing nuclear power

International Nuclear Information System (INIS)

Robinson, Colin

1990-01-01

The UK government has proposed a review of the prospects for nuclear power as the Sizewell B pressurized water reactor project nears completion in 1994. However, a delay in the completion of Sizewell B or a change of government could put off the review for some years beyond the mid 1990s. Anticipating, though, that such a review will eventually take place, issues which it should consider are addressed. Three broad categories of possible benefit claimed for nuclear power are examined. These are that nuclear power contributes to the security of energy supply, that it provides protection against long run fossil fuel price increases and that it is a means of mitigating the greenhouse effect. Arguments are presented which cost doubt over the reality of these benefits. Even if these benefits could be demonstrated, they would have to be set against the financial, health and accident costs attendant on nuclear power. It is concluded that the case may be made that nuclear power imposes net costs on society that are not justified by the net benefits conferred. Some comments are made on how a government review, if and when it takes place, should be conducted. (UK)

The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

CERN Document Server

Kelly, G N

1983-01-01

The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

Organisation of radiation protection at Sizewell Nuclear Power Plant in the UK. Report n. 290

International Nuclear Information System (INIS)

Crouail, P.; Jeannin, B.; Lefaure, C.; Panisset, L.

2004-01-01

This report first describes the organisation and management of radiation protection at Sizewell Nuclear Power Plant (UK): general organisation, organisation of the radiation protection department, goals of radiation protection at plant and corporate levels, measurement of radiation protection performance, presence of health physicists in the plant, national and international comparisons. Then, it addresses the training of workers and radiation protection specialists with respect to radiation protection, the management of zoning and surveillance (action to address the radiation risk and the contamination risk). It describes the relationships of Health physicists with contractors and other workers teams, and the relationships with safety authorities. It indicates the different outages of this organisation: general planning, information sheets, physicists work planning, reviews and meetings. It describes the management of personal dosimetry with radiation work permits and actions aimed at the reduction of doses during various operations. The last part proposes a feedback experience report and evokes the generated database, and addresses events reporting

The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

International Nuclear Information System (INIS)

Atkinson, C.A.; Collins, P.J.

1987-01-01

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

A CEGB system for the assurance of quality in the commissioning of power plant

International Nuclear Information System (INIS)

Lake, M.W.

1984-01-01

This paper outlines a system of organization, documentation and surveillance which has been fully proven in practice and looks ahead to identify the arrangements for the organization, documentation, and audit which the CEGB proposes to use in the commissioning of the Sizewell 'B' PWR, so that quality may be assured. In the UK between 1967 and 1982, the CEGB commissioned approximately 40,000 M Watts of power plant. This comprised power plants that were coal fired, oil fired, water driven, gas turbine driven and nuclear (both magnox and advanced gas cooled (AGR) reactors)

Reactor benchmarks and integral data testing and feedback into ENDF/B-VI

International Nuclear Information System (INIS)

McKnight, R.D.; Williams, M.L.

1992-01-01

The role of integral data testing and its feedback into the ENDF/B evaluated nuclear data files are reviewed. The use of the CSEWG reactor benchmarks in the data testing process is discussed and selected results based on ENDF/B Version VI data are presented. Finally, recommendations are given to improve the implementation in future integral data testing of ENDF/B

National Nuclear Corporation Limited report and financial statements 31 March 1988

International Nuclear Information System (INIS)

1988-01-01

The Group is engaged in designing and constructing power stations and nuclear power reactors and other related work and in associated research and development. Work on the Advanced Gas Cooled Reactor Power Stations at Heysham II and Torness has continued during the year. A variety of work continues to be undertaken on completed stations at Hinkley Point 'B', Heysham I, Hartlepool and Dungeness 'B' through the company's Operational Support Unit which was formed during the year. The Unit is pursuing work on other nuclear stations. Work on the Sizewell 'B' PWR under the Central Electricity Generating Board's Project Management Team is proceeding well. In June 1988 the company was awarded a contract to assist the Ministry of Defence in the management of several major projects. The accounts are presented. It is shown that the profit available to the shareholders for the year ended 31 March 1988 is Pound 4,219,000. Research and development has been undertaken on AGRs, PWRs and fast reactors. Its technologies cover chemistry, materials, heat transfer, fluid flow, mechanical engineering and instrumentation. (author)

105-B Reactor museum feasibility assessment (Phase 2) project

International Nuclear Information System (INIS)

Heckel, R. P.

2000-01-01

This 105-B Reactor Museum feasibility assessment project report documents project activities that have been performed, including a review and assessment of previously existing information, a walk-through of the facility, an assessment of potential hazards, and selection of mitigative measures deemed to be appropriate to allow unescorted access by members of the public to a specified primary tour route

IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

International Nuclear Information System (INIS)

2003-01-01

Description: B and W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (S SCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system. The general objective of the S SCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D 2 O-H 2 O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO 2 clad in stainless steel or 93%-enriched UO 2 -ThO 2 (Nth/N 15) pellets clad in aluminum. The D 2 O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO 2 fuel, and neutron age measurements in ThO 2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data. A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program. A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. A third report addresses issues that bear on the problems associated with

Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-08-15

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

Physics and nuclear power

International Nuclear Information System (INIS)

Buttery, N E

2008-01-01

Nuclear power owes its origin to physicists. Fission was demonstrated by physicists and chemists and the first nuclear reactor project was led by physicists. However as nuclear power was harnessed to produce electricity the role of the engineer became stronger. Modern nuclear power reactors bring together the skills of physicists, chemists, chemical engineers, electrical engineers, mechanical engineers and civil engineers. The paper illustrates this by considering the Sizewell B project and the role played by physicists in this. This covers not only the roles in design and analysis but in problem solving during the commissioning of first of a kind plant. Looking forward to the challenges to provide sustainable and environmentally acceptable energy sources for the future illustrates the need for a continuing synergy between physics and engineering. This will be discussed in the context of the challenges posed by Generation IV reactors

Nuclear power and public acceptance in the UK

International Nuclear Information System (INIS)

Baker, J.W.

1985-01-01

The subject is covered in sections: introduction; UK nuclear experience (experience of the Central Electricity Generating Board; impartial assessment); Sizewell inquiry (into proposal to construct a PWR based plant as Sizewell B; inquiry process; ground covered; economics and safety; project management); public acceptance (sociological studies); long-term programme. (U.K.)

German Phase B [risk study] highlights the role of [reactor] accident management

International Nuclear Information System (INIS)

Anon.

1989-01-01

Phase B of the German probabilistic risk assessment study, now scheduled for publication this month, suggests that reactor accident management measures can prevent or mitigate about 90 per cent of event sequences. (author)

Current training initiatives at Nuclear Electric plc

International Nuclear Information System (INIS)

Fowler, C.D.

1993-01-01

Nuclear Electric, one of the three generating companies to emerge from the demise of the U.K.'s Central Electricity Generating Board (CEGB), owns and operates the commercial nuclear power stations in England and Wales. The U.K. government proscribed further construction beyond Sizewell B, the United Kingdom's first pressurized water reactor (PWR) station, pending the outcome of a review of the future of nuclear power to be held in 1994. The major challenges facing Nuclear Electric at its formation in 1990 were therefore to demonstrate that nuclear power is safe, economical, and environmentally acceptable and to complete the PWR station under construction on time and within budget. A significant number of activities were started that were designed to increase output, reduce costs, and ensure that the previous excellent safety standards were maintained. A major activity was to reduce the numbers of staff employed, with a recognition from the outset that this reduction could only be achieved with a significant human resource development program. Future company staff would have to be competent in more areas and more productive. This paper summarizes some of the initiatives currently being pursued throughout the company and the progress toward ensuring that staff with the required competences are available to commission and operate the Sizewell B program in 1994

Checking out the CEGB's sums

International Nuclear Information System (INIS)

Heald, D.

1983-01-01

Accurate investment plans are difficult to make for projects such as power stations because of the timescale over which economic growth, technical performance and relative fuel prices must be estimated. For the Sizewell B PWR reactor this is 71/2 years for construction and 35 years for operating life. Past errors in forecasting, the political opposition from environmental groups and the Conservative governments dislike of nationalized industries have all combined to make the Central Electricity Generating Board (CEGB) defensive about the figures it uses to promote its power station building programme. These figures are questioned here in a summary of a report by the Electricity Consumer's Council. In particular it is suggested that the CEGB's estimate of the net effective cost was systematically biased in favour of nuclear power over coal, that the sensitivity analysis was inadequate, that the 5% discount rate chosen by the CEGB favours nuclear power with its higher capital costs and that the external presentation of results of past investments was misleading. The CEGB in its submission to the Sizewell B enquiry has rectified the first two criticisms. However, the continued use of the 5% discount rate without sensitivity tests is questioned. A table of comparative generation costs is also discussed. (U.K.)

PWR training from conception to criticality

International Nuclear Information System (INIS)

Molloy, B.

1993-01-01

Since the accident at the Three Mile Island Pressurized Water Reactor in 1979, training of reactor personnel has been critically evaluated and reviewed. In the United Kingdom, the building of the first Pressurized Water Reactor at Sizewell in Suffolk, has brought staff training into sharp relief also. This article looks at the training program, set up in response to concerns over safety, which has been evolving over the last ten years. Simulators are widely used so that staff, especially reactor operators, are thoroughly conversant with operational technology long before reactor commissioning takes place. (UK)

Nuclear economics in a nutshell

International Nuclear Information System (INIS)

Birkby, E.

1984-01-01

Two main questions are discussed from the economics viewpoint; how the Central Electricity Generating Board (CEGB) decides which of its electricity generating stations to run to meet the demand for electricity at any point in time; and how the CEGB analyse possible future investments in new generating capacity. These questions are considered generally and then with the particular case of the proposed Sizewell B PWR reactor in mind. The cost of producing electricity has three components, the capital cost, the fuel cost and operating cost. These must be considered using meaningful and comparable figures. (U.K.)

The reactor kinetics code tank: a validation against selected SPERT-1b experiments

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

Nuclear power now and in the future

Energy Technology Data Exchange (ETDEWEB)

Collier, J G [Nuclear Electric (United Kingdom)

1991-08-01

The future of the nuclear industry in the United Kingdom is considered from the perspective of the new public sector utility, Nuclear Electric, set up to retain control of nuclear power stations on the privatization of the rest of the electricity supply industry. Two major objectives are the increased nuclear generation of electricity and the cutting of costs. These are discussed in terms of life extension programmes for the magnox reactors, improved performance of AGR reactors and expectations for the Sizewell B PWR station now under construction; waste management, reactor decommissioning and fuel-cycle costs are also considered. Economic, environmental and political criteria are outlined which need to be addressed in relation to the government's review of nuclear power in 1991. Because of the marginal economic advantages of nuclear power in the United Kingdom, it will be important to quantify the environmental and diversity benefits of this source. (UK).

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Vacuum switchgear for power station auxiliary switchboards

International Nuclear Information System (INIS)

Coombs, P.E.

1992-01-01

Sizewell B is the first UK power station in which vacuum switchgear is used for the auxiliary switchboards. Previously the 3.3kV, 6.6kV or 11kV switchgear has used air-break circuit breakers and fused air-break contactors, known as motor starting devices or fused switching devices (FSD). The use of vacuum interrupters is therefore a new technology in this application, although it has been established in the UK distribution network and in industrial installations from the mid 1970s. Vacuum switchgear was already in use in the USA for power station auxiliary switchgear at the time that it was proposed for Sizewell B. The Sizewell B high voltage auxiliary switchgear comprises eight Unit and Station Auxiliary Switchboards at 3.3kV and 11kV, and four 3.3kV Essential Switchboards for the essential safety related circuits, making a total of 65 circuit breakers plus FSD panels. (Author)

Reflections on Britain's nuclear history: a conversation with Lord Hinton

International Nuclear Information System (INIS)

Lowry, D.

1984-09-01

This record of a discussion with Lord Hinton, who was in charge of the industrial side of the U.K. atomic energy project, and later Chairman of the Central Electricity Generating Board, covers the following topics: history of the development and construction of reactors for the production of plutonium and, later, nuclear power; secrecy; early relations between Central Electricity Authority (and later, CEGB), the Government, and Atomic Energy Authority; Government policy on the nuclear power programme; decision making in the area of nuclear power; economics of nuclear power; the nuclear industry; safety, insurance, National Nuclear Inspectorate; support research and development in CEGB; Sizewell-B proposed PWR. (U.K.)

The probability of containment failure by steam explosion in a PWR

International Nuclear Information System (INIS)

Briggs, A.J.

1983-12-01

The study of the risk associated with operation of a PWR includes assessment of severe accidents in which a combination of faults results in melting of the core. Probabilistic methods are used in such assessment, hence it is necessary to estimate the probability of key events. One such event is the occurrence of a large steam explosion when molten core debris slumps into the base of the reactor vessel. This report considers recent information, and recommends an upper limit to the range of probability values for containment failure by steam explosion for risk assessment for a plant such as the proposed Sizewell B station. (U.K.)

The application of the ALARP principle to the design of nuclear power stations

International Nuclear Information System (INIS)

Pugh, I.G.; Eaton, J.R.P.; Dutton, L.M.C.; Youell, F.P.

1987-01-01

A summary is given of the numerical fundamental design criteria specified to ensure nuclear reactor safety; these numerical criteria are compared to ICRP dose limits in order to demonstrate the important role of the criteria in achieving 'ALARP'. The design strategy which has been applied to both AGR and PWR power stations to ensure compliance with the design criteria and guidelines is described. The application of the strategy to Heysham II/ Torness and Sizewell 'B' demonstrates its effectiveness. Details of radiation exposures to operators and the public both in normal operation and following accidents are given including examples of cost-benefit analysis studies. (UK)

Emergency response guide-B ECCS guideline evaluation analyses for N reactor

International Nuclear Information System (INIS)

Chapman, J.C.; Callow, R.A.

1989-07-01

INEL conducted two ECCS analyses for Westinghouse Hanford. Both analyses will assist in the evaluation of proposed changes to the N Reactor Emergency Response Guide-B (ERG-B) Emergency Core System (ECCS) guideline. The analyses were a sensitivity study for reduced-ECCS flow rates and a mechanistically determined confinement steam source for a delayed-ECCS LOCA sequence. The reduced-ECCS sensitivity study established the maximum allowable reduction in ECCS flow as a function of time after core refill for a large break loss-of-coolant accident (LOCA) sequence in the N Reactor. The maximum allowable ECCS flow reduction is defined as the maximum flow reduction for which ECCS continues to provide adequate core cooling. The delayed-ECCS analysis established the liquid and steam break flows and enthalpies during the reflood of a hot core following a delayed ECCS injection LOCA sequence. A simulation of a large, hot leg manifold break with a seven-minute ECCS injection delay was used as a representative LOCA sequence. Both analyses were perform using the RELAP5/MOD2.5 transient computer code. 13 refs., 17 figs., 3 tabs

Proposed Sizewell 'B' nuclear power station

Energy Technology Data Exchange (ETDEWEB)

1981-10-01

A report by the County Planning Officer is presented. The site history and planning policies, consultations and publicity, environmental impact, safety and national energy policy are considered. Twelve appendices cover various topics, including various representations, and a report on safety by Professor D C Leslie, the Technical Adviser to the Councils. (UK).

Reactor pressure vessel steels ASTM A533B and A508 Cl.2

International Nuclear Information System (INIS)

Pelli, R.; Kemppainen, M.; Toerroenen, K.

1979-11-01

This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

The development of quality assurance and quality management in civil engineering

International Nuclear Information System (INIS)

Haste, N.D.

1991-01-01

The paper describes the development of Quality Assurance and Quality Management during construction at the Sizewell-B power plant from the lead provided in the Client's documentation. The hard lessons learnt from dealing with major suppliers and sub-contractors are explained. The paper concludes by describing the ways in which total Quality Management is now addressed within Laings Civil Engineering following the experiences at Sizewell and elsewhere. (author)

Testing ENDF/B-V data for thermal reactors

International Nuclear Information System (INIS)

Craig, D.S.

1982-10-01

Lattice parameters have been calculated for some thermal reactor benchmark lattices using ENDF/B-V data. These lattices were TRX-1, -2; BAPL-UO 2 -1,-2,-3; BNL-ThO 2 - 233 UO 2 -H 2 0-1,-2,-3; MIT-4,-5,-6; and PNL-31,-33,-35 (infinite lattices). In addition, parameters were calculated for 3 ZEEP lattices, 3 High-Conversion U0 2 -H 2 0 lattices, and 7 BNL-Th0 2 - 233 U0 2 -D 2 0 lattices. These calculations were made using the integral transport cell code RAHAB with the resonance reaction rates obtained using the OZMA code operating in the discrete ordinate mode. This code calculates the resonance rates allowing for the interaction of all resonances. Four group reaction rates for use in method comparisons are given for several lattices. The author discusses the use of the OZMA code for these calculations, including the choice of options and the orders of the angular quadratures, and compares results obtained using the CRNL thermal scattering data with those obtained using ENDF/B data

Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

International Nuclear Information System (INIS)

1995-03-01

This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

Use of PSA in a regulatory framework

International Nuclear Information System (INIS)

Ross, P.J.

1994-01-01

The paper will briefly describe the use of PSA in the licensing process for the Sizewell 'B' PWR Power Station currently under construction in the U.K. There are two distinct phases in the licensing process - (i) A PSA has been performed to support the application to construct Sizewell 'B'. At that stage the PSA was used as a design tool (along with deterministic design requirements) for Sizewell 'B' and as such lead to a number of significant design changes in the early design process. (ii) A PSA is currently being performed to support the application to operate Sizewell 'B'. The PSA is required to support the claim that the design has included all reasonably practical measures to prevent and mitigate accidents. The comprehensive PSA being produced for the second phase of the licensing process will be described. The way the regulators/designer/analysts have interacted over the years has affected the scope, complexity, detail and bias of the comprehensive PSA. The paper will discuss these issues and highlight some of the more significant ones. The benefits and drawbacks of providing a PSA in a regulatory framework will be discussed. One of the conclusions of the paper is that the use of true ''best-estimates'' in the PSA is difficult to achieve in a regulatory framework where persistent bias to the conservative side is apparent in the designers, analysts and regulators judgements. The usefulness of the PSA is therefore, potentially, compromised by giving misleading outputs or diverting resources to unnecessary areas. (author)

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

International Nuclear Information System (INIS)

Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

1993-01-01

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

The UK nuclear power industry

International Nuclear Information System (INIS)

Collier, J. G.

1995-01-01

In the United Kingdom, nuclear power plants are operated by three companies: Nuclear Electric (NE), Scottish Nuclear (SN), and British Nuclear Fuels plc (BNFL). The state-operated power industry was privatized in 1989 with the exception of nuclear power generation activities, which were made part of the newly founded (state-owned) NE and SN. At the same time, a moratorium on the construction of new nuclear power plants was agreed. Only Sizewell B, the first plant in the UK to be equipped with a pressurized water reactor, was to be completed. That unit was first synchronized with the power grid on February 14, 1995. Another decision in 1989 provided for a review to be conducted in 1994 of the future of the peaceful uses of nuclear power in the country. The results of the review were presented by the government in a white paper on May 9, 1995. Accordingly, NE and SN will be merged and privatized in 1996; the headquarters of the new holding company will be in Scotland. The review does not foresee the construction of more nuclear power plants. However, NE hopes to gain a competitive edge over other sources of primary energy as a result of this privatization, and advocates construction of a dual-unit plant identical with Sizewell B so as to avoid recurrent design and development costs. Outside the UK, the company plans to act jointly with the reactor vendor, Westinghouse, especially in the Pacific region; a bid submitted by the consortium has been shortisted by the future operator of the Lungmen nuclear power plant project in Taiwan. In upgrading the safety of nuclear power plants in Eastern Europe, the new company will be able to work through existing contacts of SN. (orig.) [de

Britain's first pressurised-water reactor

International Nuclear Information System (INIS)

Kenward, M.

1982-01-01

The recent announcement that the public inquiry into the CEGB's plans to build a PWR at Sizewell will begin in January 1983 and the statement which followed from the task force that was set up in July 1981 to consider the future of the PWR programme in the UK, are considered. The relevant time scales, costs and safety, in particular the cost incurred due to the added safety features for the British PWR, are discussed. The effect of political aspects on the future of the PWR in Britain is considered. (U.K.)

Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor-borne

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This document describes the specific dispositions relative to the nuclear reactor-borne domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor-borne and of the retraining, the Passerelle formation, are presented. (A.L.B.)

Probabilistic assessment of faults

International Nuclear Information System (INIS)

Foden, R.W.

1987-01-01

Probabilistic safety analysis (PSA) is the process by which the probability (or frequency of occurrence) of reactor fault conditions which could lead to unacceptable consequences is assessed. The basic objective of a PSA is to allow a judgement to be made as to whether or not the principal probabilistic requirement is satisfied. It also gives insights into the reliability of the plant which can be used to identify possible improvements. This is explained in the article. The scope of a PSA and the PSA performed by the National Nuclear Corporation (NNC) for the Heysham II and Torness AGRs and Sizewell-B PWR are discussed. The NNC methods for hazards, common cause failure and operator error are mentioned. (UK)

Multilink manipulator computer control: experience in development and commissioning

International Nuclear Information System (INIS)

Holt, J.E.

1988-11-01

This report describes development which has been carried out on the multilink manipulator computer control system. The system allows the manipulator to be driven using only two joysticks. The leading link is controlled and the other links follow its path into the reactor, thus avoiding any potential obstacles. The system has been fully commissioned and used with the Sizewell ''A'' reactor 2 Multilink T.V. manipulator. Experience of the use of the system is presented, together with recommendations for future improvements. (author)

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

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.)

The economic aspect

International Nuclear Information System (INIS)

MacKerron, G.

1984-01-01

The subject is covered in sections: introduction (indicating the importance of 'back end' operations in the economics of nuclear power production); irradiated fuel transport costs in the UK (the Sizewell PWR; existing UK reactors); economic appraisal; past nuclear economics; future nuclear economics; (electricity demand; technological maturity; social and political factors; competition to nuclear power). (U.K.)

Nuclear economics and privatisation

International Nuclear Information System (INIS)

Evans, N.L.

1987-01-01

In this paper the compatibility of the British government's two policies of privatisation of the electricity supply industry and the development of nuclear power is considered. The structures that a privatised electricity supply industry might have looked at, especially those which might have greatest effect on the position of nuclear power. Only three alternatives for Sizewell-B - a pressurized water reactor, an advanced gas-cooled reactor or a large coal-fired power station were considered. The economic case was made using a discount rate o 5% -lower than would be taken by private sector companies. This has an important effect on the economic advantage of one fuel over another. Changes in the electricity supply industry with privatisation are discussed. These might lead to consideration of a greater range of options for meeting demand (eg renewable energy sources) as the emphasis would be on profitability. (UK)

Nuclear privatization

International Nuclear Information System (INIS)

Jeffs, E.

1995-01-01

The United Kingdom government announced in May 1995 plans to privatize the country's two nuclear generating companies, Nuclear Electric and Scottish Nuclear. Under the plan, the two companies will become operating divisions of a unified holding company, to be called British Electric, with headquarters in Scotland. Britain's nuclear plants were left out of the initial privatization in 1989 because the government believed the financial community would be unwilling to accept the open-ended liability of decommissioning the original nine stations based on the Magnox gas-cooled reactor. Six years later, the government has found a way around this by retaining these power stations in state ownership, leaving the new nuclear company with the eight Advanced Gas-cooled Reactor (AGR) stations and the recently completed Sizewell B PWR stations. The operating Magnox stations are to be transferred to BNFL, which operates two Magnox stations of their own at Calder Hall and Chapelcross

An assessment of the radiological consequences of releases to groundwater following a core-melt accident at the Sizewell PWR

International Nuclear Information System (INIS)

Maul, P.R.

1984-03-01

In the extremely unlikely event of a degraded core accident at the proposed Sizewell PWR it is theoretically possible for the core to melt through the containment, after which activity could enter groundwater directly or as a result of subsequent leaching of the core in the ground. The radiological consequences of such an event are analysed and compared with the analysis undertaken by the NRPB for the corresponding releases to atmosphere. It is concluded that the risks associated with the groundwater route are much less important than those associated with the atmospheric route. The much longer transport times in the ground compared with those in the atmosphere enable countermeasures to be taken, if necessary, to restrict doses to members of the public to very low levels in the first few years following the accident. The entry of long-lived radionuclides into the sea over very long timescales results in the largest contribution to population doses, but these are delivered at extremely low dose rates which would be negligible compared with background exposure. (author)

Application of a new cross section library based on ENDF/B-IV to reactor core analysis

International Nuclear Information System (INIS)

Lima Bezerra, J. de.

1991-04-01

The use of the ENDF/B-IV library in the LEOPARD code for the Angra-1 reactor simulation is presented. The results are compared to those obtained using the ENDF/B-II library and show better values for the power distribution but an underestimated global reactivity as compared to experimental results. (F.E.). 1 ref, 55 figs, 1 tab

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

Enrichment reduction calculations for the DIDO reactor. App. B

International Nuclear Information System (INIS)

Constantine, G.; Javadi, M.; Thick, E.

1985-01-01

The possibility has been raised that DIDO/PLUTO type heavy water moderated reactors can be operated with fuel of lower than the 75% enrichment material currently in use with the object of increasing the proliferation resistance of the fuel cycle. This paper sets out to examine the reactor physics aspects of enrichment reductions to 45% and 20% for Harwell's MTR's as part of an IAEA collaborative exercise currently being conducted to examine the topic in a more general way for the whole class of heavy water moderated reactors. The reactor physics tool used at Harwell is WIMSE, the Winfrith Improved Multigroup Scheme, a suite of linked reactor physics codes which has been used extensively for light water, heavy water and graphite moderated thermal reactors. The course of the calculations and the WIMSE modules involved in this study are described briefly

Neutronic characteristics of linear-assembly breed-and-burn reactors

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2012-01-01

Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

National report on nuclear power plant control and instrumentation in the United Kingdom

International Nuclear Information System (INIS)

Goodings, A.

1992-01-01

This paper notes that, although no fundamental, NPPC and I technical developments have taken place in the UK since 1989, enormous organizational changes have occurred. The influence of these changes on instrumentation capability is discussed and the current situations on the Magnox, AGR and Fast Reactors and on the Sizewell PWR are described. Work on pulse-coded-logic reactor trip systems based on conventional microprocessor components is noted and a new, Dungeness B AGR, single channel trip systems which uses these principles is described. Other developments in sensors, instrumentation, ultrasonics and under-sodium viewing are also described as are studies on software reliability, human factors engineering and related topics. Comments are made on general aspects of computers in reactor safety systems. It is concluded that the last two years have seen considerable progress in the UK despite the interruptions and problems generated by organizational changes. It is felt that these changes may have produced a new sense of urgency and better insights into the needs of the industry. (author). 2 refs

UK minister and science advisors reassert roles of PWRs and FBRs

International Nuclear Information System (INIS)

Fishlock, D.

1988-01-01

Britain continues to vigorously debate the wisdom of privatizing its electricity supply system by breaking the Central Electricity Generating Board into three separate parts, with specific consequences for nuclear energy. What is already clear is that the proposed changes must have a profound impact on all long-term activities, from the future of the fast reactor to national research into the environmental impact of electricity supply. The Government stated that future use of nuclear energy is necessary. The Government then published a nuclear study by its own top technical advisors, the Advisory Council on Science and Technology (ACOST). ACOST recommends that Britain should abandon further research and development on gas-cooled reactors and concentrate wholeheartedly on the PWR, following what it says is the public inquiry's unequivocal endorsement of the Sizewell-B design. The Advisory Council also comments on the fast reactor, in the context of what it calls the Government's responsibility to underpin long-term ventures in power generation. It states that the technology is approaching commercialization

An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

2012-07-01

The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

International Nuclear Information System (INIS)

Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

1987-01-01

A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

Statements by the National Executive Committee to the 85. annual conference of the Labour Party to be held in the Empress Ballroom, Blackpool, from 29 September to 3 October, 1986

International Nuclear Information System (INIS)

1986-01-01

All the statements by the National Executive Committee to the Annual Labour Party Conference held in Blackpool in 1986 are included. As well as statements on Social Ownership, the Environment, Social Security and Taxation, Defence conversion and costs, Shipbuilding, the rights of people at work and policies on low pay, there is a 5 page section on Civil Nuclear Power. This restates the resolution passed at the 1985 conference and identifies problems with nuclear power (the risk of an accident at a nuclear power station and the dangers of disposing of nuclear waste). Labour's plan is to move away from reliance on nuclear energy to a strategy based on coal, conservation and alternative energy resources. The policy for the different reactor types is presented including opposition to a PWR reactor at Sizewell-B. The Party's policies on nuclear waste and reprocessing, and health and radiation are stated, and Labour's alternative summarized. (UK)

Optimization of a heterogeneous catalytic hydrodynamic cavitation reactor performance in decolorization of Rhodamine B: application of scrap iron sheets.

Science.gov (United States)

Basiri Parsa, Jalal; Ebrahimzadeh Zonouzian, Seyyed Alireza

2013-11-01

A low pressure pilot scale hydrodynamic cavitation (HC) reactor with 30 L volume, using fixed scrap iron sheets, as the heterogeneous catalyst, with no external source of H2O2 was devised to investigate the effects of operating parameters of the HC reactor performance. In situ generation of Fenton reagents suggested an induced advanced Fenton process (IAFP) to explain the enhancing effect of the used catalyst in the HC process. The reactor optimization was done based upon the extent of decolorization (ED) of aqueous solution of Rhodamine B (RhB). To have a perfect study on the pertinent parameters of the heterogeneous catalyzed HC reactor, the following cases as, the effects of scrap iron sheets, inlet pressure (2.4-5.8 bar), the distance between orifice plates and catalyst sheets (submerged and inline located orifice plates), back-pressure (2-6 bar), orifice plates type (4 various orifice plates), pH (2-10) and initial RhB concentration (2-14 mg L(-1)) have been investigated. The results showed that the highest cavitational yield can be obtained at pH 3 and initial dye concentration of 10 mg L(-1). Also, an increase in the inlet pressure would lead to an increase in the ED. In addition, it was found that using the deeper holes (thicker orifice plates) would lead to lower ED, and holes with larger diameter would lead to the higher ED in the same cross-sectional area, but in the same holes' diameters, higher cross-sectional area leads to the lower ED. The submerged operation mode showed a greater cavitational effects rather than the inline mode. Also, for the inline mode, the optimum value of 3 bar was obtained for the back-pressure condition in the system. Moreover, according to the analysis of changes in the UV-Vis spectra of RhB, both degradation of RhB chromophore structure and N-deethylation were occurred during the catalyzed HC process. Copyright © 2013 Elsevier B.V. All rights reserved.

The United Kingdom

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The history of nuclear development in the United Kingdom is traced, together with the factors (eg electricity demand, fossil fuel prices, political policy) that have influenced policy choices. Many of the research establishments and the fissile material production plants were established in the late 1940s. In these early years the civil and military development programmes were not separate. The initial growth period was in the 1950s when the plan was to build twelve reactors in three groups of four over ten years. The first were to be Magnox type reactors. This was trebled in 1957 and the advanced gas cooled reactor was developed. The Windscale fire in 1957 affected confidence in the nuclear programme. However, five AGR twin-reactor stations were ordered in 1964. These ran into construction problems and delay and it was not until 1980 that more new units were contemplated. The plan then was to build a pressurized water reactor at Sizewell-B. A delay to allow for the public inquiry put back the start of this until 1987 with commissioning planned for 1994. A list of the UK nuclear power stations is given and their operating performance discussed. The present situation and future prospects for nuclear power are reviewed. (U.K.)

Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

International Nuclear Information System (INIS)

Wheeler, J.K.; Pallotta, A.S.

1992-01-01

The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

Dose estimation in B16 tumour bearing mice for future irradiation in the thermal column of the TRIGA reactor after B/Gd/LDL adduct infusion

Energy Technology Data Exchange (ETDEWEB)

Protti, N., E-mail: nicoletta.protti@pv.infn.it [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy); Ballarini, F.; Bortolussi, S. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy); Bruschi, P. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy); Stella, S. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy); Geninatti, S.; Alberti, D.; Aime, S. [University of Torino, Chemistry Department, via Nizza 52, 10126 Torino (Italy); Altieri, S. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy)

2011-12-15

To test the efficacy of a new {sup 10}B-vector compound, the B/Gd/LDL adduct synthesised at Torino University, in vivo irradiations of murine tumours are in progress at the TRIGA Mark II reactor of the Pavia University. A localised B16 melanoma tumour is generated in C57BL/6 mice and subsequently infused with the adduct. During the irradiation, the mouse will be put in a shield to protect the whole body except the tumour in the back-neck area. To optimise the treatment set-up, MCNP simulations were performed. A very simplified mouse model was built using MCNP geometry capabilities, as well as the geometry of the shield made of 99% {sup 10}B enriched boric acid. A hole in the shield is foreseen in correspondence of the back-neck region. Many configurations of the shield were tested in terms of neutron flux, dose distribution and mean induced activity in the tumour region and in the radiosensitive organs of the mouse. In the final set-up, up to five mice can be treated simultaneously in the reactor thermal column and the neutron fluence in the tumour region for 10 min of irradiation is of about 5 Multiplication-Sign 10{sup 12} cm{sup -2}.

ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors

International Nuclear Information System (INIS)

Nishimura, Hideo

1977-01-01

1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC 2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma 0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma 0 are not guaranteed because of the approximations used in the unresolved resonance region

Maximization of burning and/or transmutation (B/T) capacity in coupled spectrum reactor (CSR) by fuel and core adjustment

International Nuclear Information System (INIS)

Aziz, F.; Kitamoto, Asashi.

1996-01-01

A conceptual design of burning and/or transmutation (B/T) reactor, based on a modified conventional 1150 MWe-PWR system, consisted of two core regions for thermal and fast neutrons, respectively, was proposed herein for the treatments of minor actinides (MA). In the outer region 237 Np, 241 Am, and 243 Am burned by thermal neutrons, while in the inner region 244 Cm was burned mainly by fast neutrons. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio. The maximization of B/T capacity in CSR were done by, first, increasing the radius of the inner region. Second, reducing the coolant to fuel volume ratio, and third, choosing a suitable B/T fuel type. The result of the calculations showed that the equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute up to 808 kg of MA in a single reactor core effectively and safely. (author)

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Reactor lattice codes

International Nuclear Information System (INIS)

Kulikowska, T.

2001-01-01

The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

Study of parameters affecting the conversion in a plug flow reactor for reactions of the type 2A→B

Science.gov (United States)

Beltran-Prieto, Juan Carlos; Long, Nguyen Huynh Bach Son

2018-04-01

Modeling of chemical reactors is an important tool to quantify reagent conversion, product yield and selectivity towards a specific compound and to describe the behavior of the system. Proposal of differential equations describing the mass and energy balance are among the most important steps required during the modeling process as they play a special role in the design and operation of the reactor. Parameters governing transfer of heat and mass have a strong relevance in the rate of the reaction. Understanding this information is important for the selection of reactor and operating regime. In this paper we studied the irreversible gas-phase reaction 2A→B. We model the conversion that can be achieved as function of the reactor volume and feeding temperature. Additionally, we discuss the effect of activation energy and the heat of reaction on the conversion achieved in the tubular reactor. Furthermore, we considered that dimerization occurs instantaneously in the catalytic surface to develop equations for the determination of rate of reaction per unit area of three different catalytic surface shapes. This data can be combined with information about the global rate of conversion in the reactor to improve regent conversion and yield of product.

Ship propulsion reactors technology

International Nuclear Information System (INIS)

Fribourg, Ch.

2002-01-01

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.)

Advanced gas-cooled reactors (AGR)

Energy Technology Data Exchange (ETDEWEB)

Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

1981-01-15

The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

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

Licensing of nuclear reactor operators

International Nuclear Information System (INIS)

1979-09-01

Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

Energy Technology Data Exchange (ETDEWEB)

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

Mirror fusion reactor design

International Nuclear Information System (INIS)

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

1979-01-01

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

Investment with a risk attached

International Nuclear Information System (INIS)

Barrie, C.

1996-01-01

The commercial pressures on the United Kingdom's privatised nuclear power company, British Electric, are examined. The demand to deliver increasing dividends to shareholders raises questions about the ability of the company to ensure continuing safe operation of its seven AGR power stations and the Sizewell-B PWR. Investors expectations will require that the AGR stations be kept running at a high level of productivity and that the approval of the Nuclear Installations Inspectorate be sought for life extension plans for some of these ageing reactors. Yet the City expects a further reduction in engineering manpower on top of the 18% cuts over recent years made to improve productivity in the run up to privatisation. Past cuts have inevitably led to a reduction in the company's pool of experienced personnel, so a further concern is that outside contractors brought in in the future to make up shortfalls in manpower will not have the degree of expertise necessary adequately to maintain the safe operation of the AGR reactors whose design is unique to the United Kingdom. (UK)

Materials and fabrication requirements for APWR systems

International Nuclear Information System (INIS)

Boothby, R.M.; Hippsley, C.A.; Gorton, O.K.; Garwood, S.J.

1995-01-01

Materials specifications for advanced pressurized water-cooled reactor (APWR) systems are generally based on existing designs, with improved materials and fabrication procedures being developed to counter known degradation effects. In this paper, materials ageing and degradation mechanisms in PWR primary circuit pressure boundary components (i.e. the reactor pressure vessel (RPV), control rod drive mechanisms (CRDMs), coolant piping, coolant pump casing, pressurizer, and steam generators) are reviewed. Important degradation mechanisms include irradiation embrittlement of the RPV, thermal ageing embrittlement of ferritic (e.g. the pressurizer) and cast austenitic (e.g. coolant pump casing and pipe elbows) steel components and environmentally assisted cracking of steam generator tubing and CRDM penetrations. Improved materials specifications and component design and fabrication issues affecting the integrity of the pressure boundary are discussed in the light of these materials problems. Improved fabrication procedures adopted for Sizewell B, such as the utilization of ring forgings to eliminate axial welds in the RPV and steam generator shells and the use of one-piece castings for coolant pump casings, provide a benchmark against which other APWR designs may be judged. (author)

A British standard for inspection qualification - a step on the road to European harmonization

International Nuclear Information System (INIS)

Waites, C.; Leyland, K.S.; Whittle, M.J.

1993-01-01

The safety case for the Sizewell 'B' reactor requires the ultrasonic inspection of certain components to be validated independently of those directly responsible for reactor construction and operation. This paper describes how this has been done. To illustrate the way in which validation needs to be tailored to particular requirements, an example is given involving the application of a time-of-flight detection (TOFD) technique to in-service inspection of a weld from an offshore installation. The experience gained from the application above and from other work has enabled an approach to validation to be set out in a British Standard Draft for Development. The intent is to provide a methodology within which individual validations can be developed in a consistent and coherent way. A European Methodology for Inspection Qualification also builds on this and other experience and aims to promote a common approach within Europe. The intent is to define a qualification approach to NDT which provides the necessary confidence in the test in the most cost-effective way possible. (author)

A Pebble-Bed Breed-and-Burn Reactor

International Nuclear Information System (INIS)

Greenspan, Ehud

2016-01-01

The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

A Pebble-Bed Breed-and-Burn Reactor

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

2016-03-31

The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

Research and materials irradiation reactors

International Nuclear Information System (INIS)

Ballagny, A.; Guigon, B.

2004-01-01

Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

Expected characteristics of future reactors for human beings

International Nuclear Information System (INIS)

Taketani, Kiyoaki

1992-01-01

Based on four reactor safety components (namely: a) God-given safety, b) Equipment safety, c) Quick-response safety, d) Containing safety), categorical assessment is made of various nuclear reactor concepts ranging from present existing reactors to future reactors based on innovative reactor design. In pursuit of nuclear reactor safety, ultimate characteristics of the ideal nuclear reactor are expected to coincide with those of an inherently safe reactor. A definition of 'inherently safe' has already been proposed by a committee in Japan. As a realistic and existable reactor, which is as close to the ideal reactor, a future reactor which is almost the same as a global reactor, is proposed. This global reactor must be constructable anywhere on earth and must permit easy operation and maintenance by anyone. It is also discussed to identify what behavior is expected of the global reactor under various conditions. At the same time, this future reactor which includes the global reactor, should solve a) the nuclear fuel resource issue, b) efficient utilization of fission energy and c) environmental issues as the greenhouse effect. (author). 7 refs., 2 figs

The research reactor as a tool in the master in nuclear reactors in Argentina

International Nuclear Information System (INIS)

Notari, Carla

2003-01-01

complete the Master with a seminar: Nuclear Power Plants, and a Thesis. In the frame of the academic plan, multiple activities are organized related to research reactors and also to nuclear power plants. Since the very beginning the performance of selected experiments in a nuclear reactor was recognized as an extraordinary tool to give the students an insight in the principal phenomena associated with the chain reaction and the related engineering problems. This experiments have an intrinsic elevated cost, associated with the relevance of the installation and with the specialized personnel involved. CNEA provides the career with this educational instrument through the Ra-1 and RA-3 reactors located at Constituyentes and Ezeiza Atomic Centers respectively. Various activities are under way but the most established, in the Reactor Physics Course, is the estimation of kinetic parameters in RA-1 reactor. The practice includes three different experiments: Approach to critical and calibration of control rods by the compensation method: Starting in a subcritical state with source the calibration of control rod B1 vs B2 is done by introduction of the first and withdrawal of the second. The methods used are based on the Point Kinetic Model; Measurement of control rods effectivity by the rod-drop method: Separate Rod Drop of rods B1 B2 B3 of the overall ensemble B1 B2 B3 B4 and total scram starting with three withdrawn and one partially inserted, is the procedure followed to estimate the reactivity worth of B1 B2 B3 and scram. The Point Kinetic Model and the Modal Kinetic Model are used; Reactor noise technique for the estimation of reactor parameters: α and Λ. The kinetic parameters are estimated assuring that the Point Kinetic Model is valid (detection chambers near to the core), that the fluctuation of the fission density is the dominant source of the correlated part of neutron noise (measurement at low power, <10kw), the dominance of the fundamental armonic (simultaneous use of

The experimental and technological developments reactor

International Nuclear Information System (INIS)

Carbonnier, J.L.

2003-01-01

THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

Concept on coupled spectrum B/T (burning and/or transmutation) reactor for treatment of minor actinides by thermal and fast neutrons

International Nuclear Information System (INIS)

Aziz, Ferhat; Kitamoto, Asashi

1996-01-01

A conceptual design of B/T (burning and/or transmutation) reactor based on a modified conventional 1150 MWe-PWR system, with core consisted of two concentric regions for thermal and fast neutrons, was proposed herein for B/T treatment of MA (minor actinides). The B/T fuel considered was supposed such that MA discharged from 1 GWe-LWR was blended homogeneously with the composition of LWR fuel. In the outer region 23- Np, 241 Am and 243 Am were loaded and burned by thermal neutron, while in the inner region 244 Cm was loaded and burned mainly by fast neutron. The geometry of B/T fuel and the fuel assembly in the outer region was left in the same condition to those of standard PWR while in the inner region the B/T fuel was arranged in the hexagonal geometry, allowed high fuel to coolant volume ratio (V m /V f ), to keep the harder neutron spectrum. Two cases of the Coupled Spectrum B/T Reactor (CSR) with different (V m 1 f ) ratio in the inner region were studied, and the results for the tight lattice with (V m /V f ) = 0.5 showed that those isotopes approached the equilibrium composition after about 5 recycle period, when the CSR was operated under the reactivity swing of 2.8 % dk/k. The evaluations on the void coefficient of reactivity, the Doppler effect and the reactivity swing showed that the CSR concept has the inherent safety and can burn and/or transmute all kind of MA in a single reactor. This CSR can burn about 808 kg of MA in one recycle period of 3 years, which is equivalent to the discharged fuel from about 12 units of LWR in a year. (author)

Preparing the construction of a school reactor

International Nuclear Information System (INIS)

Matejka, K.

1977-01-01

The possibilities are discussed of teaching and training nuclear reactor operation and control, teaching experimental reactor physics and investigating reactor lattice parameters using a training reactor to be installed at the Faculty of Nuclear Science and Physical Engineering in Prague. Requirements are indicated for the reactor's technical design and the Faculty's possibilities to contribute to its construction. (J.B.)

When the engineering has to stop

International Nuclear Information System (INIS)

Baker, J.W.

1985-01-01

The paper proposes that, in the UK, there needs to be a fundamental change in the attitudes of engineers. Engineering alone is not enough for engineers. They need to see themselves in the broader context of society as a whole. At present there is a general distrust by the general public of science and engineering, mainly because engineers have kept quiet for too long, and have been letting facts speak for themselves when they should have been explaining them. Now it is up to the engineering profession to educate the general public in terms that they can understand, in an attempt to win back wider public support. The paper includes a discussion of some aspects of the UK nuclear power programme: the choice of reactor types; reprocessing and waste disposal; Sizewell-B enquiry; public opinion on civil nuclear power; CEGB public relations. (author)

Institutional adaptation in the UK nuclear technology industry

International Nuclear Information System (INIS)

Kemp, R.

1986-01-01

Public inquiries into controversial development proposals have themselves become increasing controversial in recent years. Despite their increasing length and cost, such proceedings remain part of the administrative process - the Inspector's report serving to 'inform the minister's mind'. This paper examines certain features of the Sizewell B Inquiry and draws attention to the ability of the applicants to respond to criticisms raised at the hearings through various forms of institutional adaptation. Changes to the Central Electricity Generating Board's case on the organization of project management, to the management of the nuclear site licensing process, and to provisions for future nuclear reactor choice are highlighted. Significantly, each of these areas had been the subject of particular investigations undertaken at the behest of the inquiry Inspector. The paper concludes by raising a number of questions concerning the implications of this singularly interactive process. (author)

A regulatory view of nuclear containment on UK licensed sites

International Nuclear Information System (INIS)

Bradford, P.M.; McNair, I.J.

1997-01-01

Members of the UK regulatory body, HM Nuclear Installations Inspectorate (NII) have previously presented conference papers and official reports which have dealt separately with either reactor applications or chemical plant applications. The objective of this paper is to draw together a brief overview of the role of containment in protecting against potential radiological and related hazards, and to describe the factors which influence the NII's assessment of containment safety cases. It draws upon the NII's experience of regulating many types of nuclear facility, from those designed in the late 1940s through to the modern plants, such as Sizewell 'B' and THORP. The paper reviews the legislative and regulatory background within which the facilities exist and are operated. Finally, the paper reviews recent, ongoing and planned research in the field of containment, which has been designed to behave under challenge. (author)

The crumbling case for nuclear power

International Nuclear Information System (INIS)

Bunyard, P.

1983-01-01

In connection with the Public Inquiry into the CEGB proposal to build a pressurised water reactor at Sizewell in Suffolk, the case for nuclear power is examined under the headings: the economics of nuclear power - how they would like them to be; systems analysis - net effective cost; CEGB prejudices the results (comparison with coal-fired plants; forecasting on various assumptions); discounting future costs; back-end costs soar (reprocessing); real reprocessing costs; AGR costs balloon. (U.K.)

Structural integrity of water reactor pressure boundary components

International Nuclear Information System (INIS)

Loss, F.J.

1977-01-01

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

Manufacture of components for Canadian reactor programs

International Nuclear Information System (INIS)

Perry, L.P.

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

Cleanup Verification Package for the 118-B-1, 105-B Solid Waste Burial Ground

International Nuclear Information System (INIS)

Capron, J.M.

2008-01-01

This cleanup verification package documents completion of remedial action, sampling activities, and compliance criteria for the 118-B-1, 105-B Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-B Reactor and P-10 Tritium Separation Project and also received waste from the 105-N Reactor. The burial ground received reactor hardware, process piping and tubing, fuel spacers, glassware, electrical components, tritium process wastes, soft wastes and other miscellaneous debris

Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-10-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

Edwards, R.M.; Power, M.A.; Bryan, M.

1992-01-01

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

Babcock and Wilcox model for predicting in-reactor densification

International Nuclear Information System (INIS)

Buescher, B.J.; Pegram, J.W.

1975-06-01

The B and W fuel densification model is used to describe the extent and kinetics of in-reactor densification in B and W production fuel. The model and approach are qualified against an extensive data base available through B and W's participation in the EEI Fuel Densification Program. Out-of-reactor resintering tests on representative pellets from each batch of fuel are used to provide input parameters to the B and W densification model. The B and W densification model predicts in-reactor densification very accurately for pellets operated at heat rates above 5 kW/ft and with considerable conservation for pellets operated at heat rates less than 5 kW/ft. This model represents a technically rigorous and conservative basis for predicting the extent and kinetics of in-reactor densification. 9 references. (U.S.)

The safety of light water reactors

International Nuclear Information System (INIS)

Pershagen, B.

1986-04-01

The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

History of 100-B Area

International Nuclear Information System (INIS)

Wahlen, R.K.

1989-10-01

The initial three production reactors and their support facilities were designated as the 100-B, 100-D, and 100-F areas. In subsequent years, six additional plutonium-producing reactors were constructed and operated at the Hanford Site. Among them was one dual-purpose reactor (100-N) designed to supply steam for the production of electricity as a by-product. Figure 1 pinpoints the location of each of the nine Hanford Site reactors along the Columbia River. This report documents a brief description of the 105-B reactor, support facilities, and significant events that are considered to be of historical interest. 21 figs

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Opening stages of the marathon at the Maltings

International Nuclear Information System (INIS)

Anon.

1983-01-01

In this report on the opening stages of the public inquiry into plans to build a new nuclear power station at Sizewell in Suffolk based on the pressurised water reactor system, the opening address of Lord Silsoe QC and the evidence of J. Baker, R. Mathews and J. Wright, all appearing for the CEGB, are summarized. Arguments examined included, economy, siting, safety, diversification of fuel supplies, and the question of using renewable energy sources for alternative power generation. (U.K.)

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Painter, S.L.

1992-12-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

DELIGHT-B/REDEL, point reactivity burnup code for high-temperature gas-cooled reactor cells

International Nuclear Information System (INIS)

Shindo, Ryuiti; Watanabe, Takashi.

1977-03-01

Code DELIGHT-2 was previously developed to analyze cell burnup characteristics and to produce few-group constants for core burnup calculation in high-temperature gas-cooled reactors. In the code, burnup dependency of the burnable poison, boron-10, is considered with the homogeneous model of space. In actuality, however, the burnable poison is used as homogeneous rods or uniform rods of small granular poison and graphite, to control the reactivity and power distribution. Precise analysis of the burnup characteristics is thus difficult because of the heterogeneity due to the configuration of poison rods. In cell burnup calculation, the DELIGHT-B, which is a modification of DELIGHT-2, takes into consideration this heterogeneous effect. The auxiliary code REDEL, a reduction of DELIGHT-B, used in combination with 3 dimensional diffusion code CITATION, is for core burnup calculation with the macro-scopic cross section model. (auth.)

Data for use in UKAEA PWR plant studies

International Nuclear Information System (INIS)

Kinnersly, S.R.; Richards, C.G.; O'Mahoney, R.

1983-05-01

Plant data represented by the RETRAN, RELAP4 and TRAC models used at Winfrith for studies of pressurised faults and small and large break loss of coolant accidents for the UK PWR are presented together with comparable data for the Sizewell B design taken from the Pre-Construction Safety Report (PCSR). The main components of the plant are described, and modelling issues, which may affect the interpretation and assessment of the data, and the historical development and use of the models, are outlined. The bulk of the report consists of tables of data with supporting figures and text for all the main items of plant modelled in the Winfrith accident studies. The data presented should be adequate to allow assessments of the Winfrith models and results to be carried out and provide a firm basis for the development of models more representative of the Sizewell B PCSR design. (U.K.)

Concept and optimization of burning and transmutation reactor in nuclear fuel recycle system

International Nuclear Information System (INIS)

Marsodi; Mulyanto; Kitamoto, Asashi.

1994-01-01

Basic concept of B/T reactor, not only produces thermal energy but also performs burning and/or transmutation of MA and long-lived FPs, was introduced here based on numerical computation model. The advantage of nuclear reaction by thermal or fast neutron was combined conceptually with each other in order to maximize the overall B/T rate obtained by a composite system of fast and thermal reactor. According to the mass balance analysis of B/T reactors with P-T treatment, fast reactor hardened neutron energy may be effective for MA burning. Furthermore, a high flux reactor operated by fast or thermal neutron could be different from a reactor with high B/T rate or high capacity for loading of MA and/or long-lived FPs. The purpose of this study is to make clear the concept and the performance of fast and thermal B/T reactor designed under high neutron utilization for HLW disposal. (author)

Evacuation decision-making at three mile island

International Nuclear Information System (INIS)

Zeigler, Donald. J.; Johnson, James. H.

1987-01-01

The accident at the Three Mile Island nuclear power plant in 1979 provoked an unanticipated and unprecedented spontaneous evacuation of people living in the area. Following the accident, revised and upgraded emergency preparedness and response regulations were issued by the Nuclear Regulatory Commission (NRC) and the Federal Emergency Management Agency. (FEMA). This includes the assumption that public education and awareness will minimise the tendency of people to evacuate spontaneously from the vicinity of an accident. This assumption is challenged. Results of an empirical test of a casual model of emergency evacuation decision-making are given. This test was devised to aid understanding of the public behaviour at the time of the Three Mile Island incident. The emergency plans for the Sizewell-B reactor are subject to brief critical consideration. It is concluded that evacuation plans need to reflect people's natural inclinations to move away from a nuclear hazard. (UK)

EBR-II Reactor Physics Benchmark Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

2017-12-28

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

Energy Technology Data Exchange (ETDEWEB)

Barral, J C [Electricite de France (EDF), 75 - Paris (France); Zaetta, A [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G [CEA/Saclay, DEN, 91 - Gif sur Yvette (France); and others

2000-07-01

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.)

Software functions for safe operation - learning from Sizewell-B

International Nuclear Information System (INIS)

Welbourne, D.

1996-01-01

Future nuclear plants will use computer-based systems extensively. Regulatory acceptance must be planned and not underestimated. Commercial software packages will simplify it, but costly analysis and demonstration may be needed. Multiplexed control needs preparation of extensive configuration data and careful checking. On-screen soft control will need consideration of the integrity of the control path. Display design should follow human factors analysis of the operators' needs, and display layout needs great care for clarity. Computer-based system with planned quality will then bring great benefits in safe operation. (author) 1 fig., 3 refs

End of the nuclear dream. The UKAEA and its role in nuclear research and development

Energy Technology Data Exchange (ETDEWEB)

Flood, M

1988-01-01

The study sets out to assess the performance of the United Kingdom Atomic Energy Authority (UKAEA) and to make recommendations for its role in the future within the overall energy needs of the United Kingdom. It assumes that the Magnox stations will be phased out after 20-25 years service, that the Advanced Gas-Cooled reactors will be shut down after 20-25 years service and no more built and that perhaps five or so Pressurized Water reactors will be built by the year 2000, including the Sizewell-B station. The first part of the report traces the institutional changes that have taken place in the UKAEA from its formation in 1954 up to the late 1980s. The current structure of the UKAEA is described in part 2. Questions of accountability are raised and its dual role of assessor and promoter of nuclear power is questioned. Part three looks at the UKAEA's role in developing nuclear power, especially the economic implications. In part four fundamental errors of judgement by the UKAEA are blamed for the failure of the UK nuclear industry to provide cheap, abundant, electricity. Practical recommendations for a major restructuring of the Authority are made.

Hinkley Point 'C' power station public inquiry: proof of evidence on emergency planning

International Nuclear Information System (INIS)

Western, D.J.

1988-09-01

A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom, adjacent to an existing nuclear power station incorporating Magnox and Advanced Gas Cooled reactors. The CEGB evidence to the Inquiry presented here introduces the concept of the Reference Accident as the basis for emergency arrangements. The description which follows of the emergency arrangements at the Hinkley Point site include: the respective responsibilities and their co-ordination of bodies such as the CEGB, external emergency services and government departments; the site emergency organization; practical aspects of the emergency arrangements; and consideration of the extension of the arrangements to a PWR on the same site. Recent developments in emergency planning, such as those arising out of post Chernobyl reviews and the Sizewell ''B'' PWR Inquiry, are taken into account. The conclusion is reached that soundly based emergency arrangements already exist at Hinkley Point which would require relatively minor changes should the proposed PWR be constructed. (UK)

Babcock experience of automated ultrasonic non-destructive testing of PWR pressure vessels during manufacture

International Nuclear Information System (INIS)

Dikstra, B.J.; Farley, J.M.; Scruton, G.

1990-01-01

Major developments in ultrasonic techniques, equipment and systems for automated inspection have lead, over a period of about ten years, to the regular application of sophisticated computer-controlled systems during the manufacture of nuclear reactor pressure vessels. Ten years ago the use of procedures defined in a code such as ASME XI might have been considered sufficient, but it is now necessary, as was demonstrated by the results of the UKAEA defect detection trials and the PISC II trials, to apply more comprehensive arrays of probes and higher test sensitivities. The ultrasonic techniques selected are demonstrated to be adequate by modelling or test-block exercises, the automated systems applied are subject to stringent quality assurance testing, and very rigorous inspection procedures are used in conjunction with a high degree of automation to ensure reproducibility of inspection quality. The state-of-the-art in automated ultrasonic testing of pressure vessels by Babcock is described. Current developments by the company, including automated flaw recognition, integrated modelling of inspection capability, and the use of electronically scanned variable-angle probes are reviewed. Examples quoted include the automated ultrasonic inspections of the Sizewell B pressurized water reactor vessel. (author)

A study into the consequences of a nuclear accident

International Nuclear Information System (INIS)

Arnott, D.G.

1987-07-01

The nuclear industry in Britain would like to believe, and would like the general public to believe, that major accidents such as that at Chernobyl in 1986, could no happen in Britain, because the design and operating procedure have been made as safe as possible. However, because the designers and operators are human, they can make mistakes. Some of these are mentioned; errors of design, errors of maintenance or inspection and errors of judgement. In spite of protestations to the contrary, a major accident could occur at Sizewell-B reactor. Given that this a real possibility, plans should be drawn up to prepare for the situation. The study considers the possible consequences of a nuclear accident under the headings, human error, how nuclear fission works, radioactivity, the truth about Chernobyl, what patterns of reactor accident are possible, what can be done (this includes meteorological information, the issuing of potassium iodate tables, radiation monitoring and evacuation). Practical issues which should concern the local authorities, especially Wrekin Council, are discussed and a recommendation made for an environmental protection officer to be appointed to keep the matter under continuing review. (U.K.)

Chooz B

International Nuclear Information System (INIS)

Barillot, Pascale; Baize, Jean-Marc

1997-01-01

This EDF press communique give information related to the exploitation of the Chooz B NPP. A calendar of the Chooz B1 and B2 NPPs exploitation is given as well as information about the local economic impact. The exploitation of the PWR reactors of the French nuclear sector corresponds to a accumulated experience of 600 year-reactor. Significant technological evolution has been recorded, namely in the test-control system, the turbo-alternator group 'Arabelle' and in the vapor generators. The reactor safety is based on the high professionalism of the exploitation personnel, on the computer-assisted behaviour allowing the choice of operators and on the conception based upon the experience accumulated by the French nuclear power plants, equivalent to 600 year-reactor operation. EDF operates a system of continual surveillance which allows the monitoring the environmental effects caused by the NPP exploitation. The following issues concerning the environment impact are reported in this document: - The effluent releases in the environment; - Health studies conducted in the NPPs' neighbourhood; - New authorizations for waste release; - Radioactive waste management. The report also mentions the French-Belgian partnership in the PWR construction, the socio-economic regional impact of the EDF activities related with the Chooz NPP operation, and the partnership with the associated service companies. Six appendices are attached to the report containing the following information: - A general layout of Chooz NPP; - Chooz B key figures; Chooz B key data; - Security and public information; - Evolution of PWR system in France; - World's nuclear systems

100-B area technical baseline report

International Nuclear Information System (INIS)

Carpenter, R.W.

1994-01-01

This document supports the environmental remediation effort of the 100-B Area by providing remediation planners with key data that characterize the 100-B and 100-C Reactor sites. It provides operational histories of the 100-B and 100-C Reactors and each of their associated liquid and solid waste sites

100-B area technical baseline report

Energy Technology Data Exchange (ETDEWEB)

Carpenter, R.W.

1994-09-01

This document supports the environmental remediation effort of the 100-B Area by providing remediation planners with key data that characterize the 100-B and 100-C Reactor sites. It provides operational histories of the 100-B and 100-C Reactors and each of their associated liquid and solid waste sites.

Safety and core design of large liquid-metal cooled fast breeder reactors

Science.gov (United States)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Analysis of the interim safe storage of reactors at the Hanford site

International Nuclear Information System (INIS)

Wang Hailiang

2014-01-01

The nine production reactors, i.e. B, C, D, DR, F, H, KE, KW and N, at the Hanford site are all water-cooled and graphite-moderated reactors with natural uranium fuel. In 1993, the U.S. Department of Energy (DOE) decided to put eight production reactors (except for B) into Interim Safe Storage (ISS) for 75 years followed by deferred one-piece removal. Reactor B will remain as a national historical landmark. By the end of 2013, six reactors C, F, D, DR, H and N had been successfully put into the ISS. Reactors KE and KW will be put into the ISS in the coming years. Taking reactor C as an example, this paper mainly talks about how to put the production reactors in the Interim Safe Storage, e.g. how to make site preparation, how to construct the safe storage enclosure (SSE) and how to perform surveillance and maintenance during the ISS period, etc. (authors)

Sewage disposal using anaerobic membrane reactor. Kenkiseimaku reactor ni yoru gesui shori

Energy Technology Data Exchange (ETDEWEB)

Fujita, Y. (Dic-Degremont Co. Ltd., Tokyo (Japan))

1991-11-01

Discussions were given on a small-scale sewage disposal of about bod 200 mg/l, for which no many examples of use have been hitherto available, using a system combining an anaerobic reactor and membrane modules. Experiments had been carried out from 1988 through 1990 as a part of the Aqua-Renaissance Project. The test equipment wza installed in the premises of the Chigasaki Coastal Research Facilities operated by the Ministry of International Trade and Industry, which used sewage flowing from the adjoining sewage treatment plant for the southern area of the Fujisawa City. The test facility consisted of a system comprising a pretreatment facility, SS decomposing reactor, fluid-bed reactor, separation membrane modules, nitrogen removing facility and micro-organism activity measurement. The test facility was constucted assuming a treatment of 10 m{sup 3} a day. The system was divided into a composite system, A system and B system to operate the system in simplified flows. As a result of comparing the composite system, A system and B system, it was found that B system can deal with wider range of disposal for a small-scale sewage treatment of about 1000 m{sup 3} a day. 6 refs., 14 figs., 3 tabs.

Technical expertise in the decision process - some comparative observations

International Nuclear Information System (INIS)

Wynne, B.

1987-01-01

The Sizewell-B public Inquiry aimed to be a 'full, fair and thorough' examination of all the facts. This could not really be achieved because the economic and safety cases cannot be exhaustively and definitively analysed. Technical examination cannot resolve these issues - only political commitment and history can, in time. This problem of uncertainties in the decision making process is discussed in detail, indicating the difficulties with large-scale public inquiries such as that for Sizewell-B. Analysis of many case studies of decision making indicates two generalisations. The first is that commitments made on the basis of values or political interests are represented as if they were the inevitable result of economic logic or scientific law. The second is that expert disagreement over technical claims is widely believed to be caused by lack of vigour, competance or impartiality on one side or the other. Recommendations for improving the decision-making process are made. (U.K.)

Compliance of the Savannah River Plant P-Reactor cooling system with environmental regulations. Demonstrations in accordance with Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972

International Nuclear Information System (INIS)

Wilde, E.W.

1985-12-01

This document presents demonstrations under Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972 for the P-Reactor cooling system at the Savannah River Plant (SRP). The demonstrations were mandated when the National Pollution Discharge Elimination System (NPDES) permit for the SRP was renewed and the compliance point for meeting South Carolina Class B water quality criteria in the P-Reactor cooling system was moved from below Par Pond to the reactor cooling water outfall, No. P-109. Extensive operating, environmental, and biological data, covering most of the current P-Reactor cooling system history from 1958 to the present are discussed. No significant adverse effects were attributed to the thermal effluent discharged to Par Pond or the pumping of cooling water from Par Pond to P Reactor. It was conluded that Par Pond, the principal reservoir in the cooling system for P Reactor, contains balanced indigenous biological communities that meet all criteria commonly used in defining such communities. Par Pond compares favorably with all types of reservoirs in South Carolina and with cooling lakes and reservoirs throughout the southeast in terms of balanced communities of phytoplankton, macrophytes, zooplankton, macroinvertebrates, fish, and other vertebrate wildlife. The report provides the basis for negotiations between the South Carolina Department of Health and Environmental Control (SCDHEC) and the Department of Energy - Savannah River (DOE-SR) to identify a mixing zone which would relocate the present compliance point for Class B water quality criteria for the P-Reactor cooling system

CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

1982-01-01

Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

Development of 'low activation superconducting wire' for an advanced fusion reactor

International Nuclear Information System (INIS)

Hishinuma, Y.; Yamada, S.; Sagara, A.; Kikuchi, A.; Takeuchi, T.; Matsuda, K.; Taniguchi, H.

2011-01-01

In the D-T burning plasma reactor beyond ITER such as DEMO and fusion power plants assuming the steady-state and long time operation, it will be necessary to consider carefully induced radioactivity and neutron irradiation properties on the all components for fusion reactors. The decay time of the induced radioactivity can control the schedule and scenarios of the maintenance and shutdown on the fusion reactor. V 3 Ga and MgB 2 compound have shorter decay time within 1 years and they will be desirable as a candidate material to realize 'low activation and high magnetic field superconducting magnet' for advanced fusion reactor. However, it is well known that J c -B properties of V 3 Ga and MgB 2 wires are lower than that of the Nb-based A15 compound wires, so the J c -B enhancements on the V 3 Ga and MgB 2 wires are required in order to apply for an advanced fusion reactor. We approached and succeeded to developing the new process in order to improve J c properties of V 3 Ga and MgB 2 wires. In this paper, the recent activities for the J c improvements and detailed new process in V 3 Ga and MgB 2 wires are investigated. (author)

Study of enzymatic reactors with microencapsulated lipase. Doctoral thesis. Estudo de reactores enzimaticos com lipase microencapsulada

Energy Technology Data Exchange (ETDEWEB)

de Franca Teixeira dos Prazeres, D.M.

1992-10-01

The work reports the development of a membrane reactor for the hydrolysis of triglycerides catalyzed by lipase B from Chromobacterium viscosum in AOT/isooctane reversed miceller system. In a preliminary phase the potential of the organic system was evaluated with comparative studies on the activity and stability of lipase B in aqueous media (emulsion) and in the alternative miceller media. A tubular ceramic membrane reactor with recirculation was selected for the olive oil hydrolysis in a reversed miceller system. The influence of the hydration degree, recirculation rate, AOT, olive oil and lipase concentration in the operation of the reactor were investigated in a batch mode. The hydration degree was identified as a critical parameter due to its influence in the separation process and in the kinetics of the reaction.

Netherlands Reactor Centre

International Nuclear Information System (INIS)

Anon.

1976-01-01

Briefly reviews the last year's work of the twenty year old Netherlands Reactor Centre (RCN) in the fields of reactor safety, fissile material, nuclear fission, non-nuclear energy systems and overseas co-operation. The annual report thus summarised is the last one to appear under the name of RCN. The terms of reference of the organisation having been broadened to include research into energy supply in general, it is to be known in future as the Netherlands Energy Research Centre (ECN). (D.J.B.)

An estimate of conditioned waste arisings to the years 2000 and 2010 for the Department of Energy power generation scenarios

International Nuclear Information System (INIS)

Fairclough, M.P.; Moore, D.C.; Tymons, B.J.

1984-09-01

An estimate of conditioned waste arisings to the years 2000 and 2010 has been made using evidence presented at the Sizewell 'B' public enquiry. The method of calculation has been based on the rate of arisings per GWe year and the power programmes of the Department of Energy. (author)

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

International Nuclear Information System (INIS)

Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

2015-01-01

Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

Digital control of research reactors

International Nuclear Information System (INIS)

Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

1991-01-01

Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

Reactivity and neutron flux measurements in IPEN/MB-01 reactor with B4C burnable poison

International Nuclear Information System (INIS)

Fer, Nelson Custodio; Moreira, Joao Manoel Losada

2000-01-01

Burnable poison rods, made of B 4 C- Al 2 O 3 pellets with 5.01 mg/cm 3 10 B concentration, have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. Several core parameters which are affected by the burnable poisons rods have been measured. The principal results, for the situation in which the burnable poison rods are located near the absorber rods of a control rod, are they cause a 29% rod worth shadowing, a reduction of 39% in the local void coefficient of reactivity, a reduction of 4.8% in the isothermal temperature coefficient of reactivity, and a reduction of 9% in the thermal neutron flux in the region where the burnable poison rods are located. These experimental results will be used for the validation of burnable poison calculation methods in the CTMSP. (author)

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

International Nuclear Information System (INIS)

1973-01-01

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

Certifying the decommissioned Shippingport reactor vessel for transport

International Nuclear Information System (INIS)

Towell, R.H.

1990-01-01

The decommissioned Shippingport reactor pressure vessel with its concentric neutron shield tank was shipped to Hanford, WA as part of the effort to restore the Shippingport Station to its original condition. The metal walls of the reactor vessel had become radioactive from neutron bombardment while the reactor was operating so it had to be shipped under the regulations for transporting radioactive material. Because of the large amount of radioactivity in the walls, 16,467 Curies, and because the potentially dispersible corrosion layer on the inner walls of both tanks was also radioactive, the Shippingport reactor vessel was transported under the most stringent of the regulations, those for a type B package. Compliance with the packaging regulations was confirmed via independent analysis by the staff of the Department of Energy certifying official and the Shippingport reactor vessel was shipped under DOE Certificate of Compliance USA/9515/B(U)

Permanent seal ring for a nuclear reactor cavity

International Nuclear Information System (INIS)

Hankinson, M.F.; Marshall, J.R.

1988-01-01

A nuclear reactor containment arrangement is described including: a. a reactor vessel which thermally expands and contracts during cyclic operation of the reactor and which has a peripheral wall; b. a containment wall spaced apart from and surrounding the peripheral wall of the reactor vessel and defining an annular thermal expansion gap therebetween for accommodating thermal expansion; and c. an annular ring seal which sealingly engages and is affixed to and extends between the peripheral wall of the reactor vessel and the containment wall

Overview of plant specific source terms and their impact on risk

International Nuclear Information System (INIS)

Desaedeleer, G.

2004-01-01

Probabilistic risk assesment and safety assessment focuses on systems and measures to prevent core meltdown, and it integrates many aspects of design and operation. It provides mapping of initiating event, frequencies onto plant damage state and through plant systems analysis, utilizes fault tree and event tree logic models, may include 'external event' analysis such as fire, flood, wind, seismic events. Percent contribution of sequences to the core damage frequency are shown for the following plants, taken as examples ZION, EDISON, OCONEE 3, SEABROOK, SIZEWELL B, MILLSTONE 3, RINGHALS 2. The presentation includes comparison of the following initiating event frequencies: loss of off-site power; small LOCA; large LOCA, steam generator tube rupture; loss of feedwater; turbine trip; reactor trip. Consequence analysis deals with: dispersion and depletion of radioactivity in the atmosphere, health effects, factors in the off-site emergency plan analyzed with codes that address the weather conditions; provision of mapping of source terms; risk diagram for early fatalities and for latent cancer fatalities

A survey on the development of advanced instrumentation and control system in NPP

International Nuclear Information System (INIS)

Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan

1993-12-01

Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: 1) N4 of EDF in France, 2) AP 600 of Westinghouse in USA, 3) NUPLEX-80+ of ABB-CE in USA, 4) CANDU in Canada, 5) Ohi 3 and 4, APWR and ABWR in Japan, 6) Belt-D in Germany, 7) Sizewell B in Britain, 8) Halden Reactor Projector in Norway, 9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author)

A survey on the development of advanced instrumentation and control system in NPP

Energy Technology Data Exchange (ETDEWEB)

Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-12-01

Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: (1) N4 of EDF in France, (2) AP 600 of Westinghouse in USA, (3) NUPLEX-80+ of ABB-CE in USA, (4) CANDU in Canada, (5) Ohi 3 and 4, APWR and ABWR in Japan, (6) Belt-D in Germany, (7) Sizewell B in Britain, (8) Halden Reactor Projector in Norway, (9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author).

Risk assessment - black art or science?

International Nuclear Information System (INIS)

Moore, G.

1988-01-01

Measures of risk can be divided into two categories, those that observe or calculate the risk of a process or project, and those that rely on the level of risk as perceived by the people during the assessment. Collection of data of accidents (where cause and effect are obvious) and experiments on animals which can then be extrapolated to humans, are two ways of risk assessment. Mathematical models and computerized simulations, using either fault tree analysis or Monte Carlo methods are explained simply. Using these methods, experts are able to perceive risk fairly realistically. However, the general public's perception of risk is often quite different, as potential risk is assessed in different ways. The concept of tolerable risk is considered, particularly with reference to nuclear reactors such as Sizewell-B. The need to inform the public of safeguards and safety procedures so they have a better understanding of the risks of nuclear power is stressed. (U.K.)

Great Britain 1994: Starting a new era for nuclear power

International Nuclear Information System (INIS)

Collier, J.G.

1994-01-01

In the course of the British program of privatization of the state-owned electricity generating industry the nuclear power plants in England and Wales were incorporated into the newly founded (and state-owned) NE in 1989. In addition, the government decided in favor of a moratorium in the construction of more nuclear power plants. Only Sizewell B, Britain's first pressurized water reactor (PWR), was to be completed. At that point in time, three other PWRs were in the planning stage. The government decision included a review of the future of nuclear technology to be held in 1994. At the beginning of 1994, the nuclear future of Britain appears in a much more positive light than had been thought possible in 1989. When founding NE, the British government endowed it with a startup capital and also instituted a tax levied on fossil energy sources, the revenue from which is being made available to NE in compensation for the old liabilities NE had to assume. (orig./UA) [de

History of research reactor fuel fabrication at Babcock and Wilcox

International Nuclear Information System (INIS)

Freim, James B.

1983-01-01

B and W Research Reactor Fuel Element facility at Lynchburg, Virginia now produces national laboratory and university fuel assemblies. The Company's 201000 square foot facility is devoted entirely to supplying research fuel and related products. B and W re-entered the research reactor fuel market in 1981

International working group on gas-cooled reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

1981-01-15

The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

Localized corrosion problems in water reactors

International Nuclear Information System (INIS)

Coriou, Henri.

1977-01-01

Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

Tritium resources available for fusion reactors

Science.gov (United States)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

OECD Halden reactor project

International Nuclear Information System (INIS)

1978-01-01

This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

Trench reactor: an overview

International Nuclear Information System (INIS)

Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

1988-01-01

Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

The benefits of nuclear power to the United Kingdom

International Nuclear Information System (INIS)

Allen, A.M.

1985-01-01

The larger of Great Britain's generating boards, the CEGB, is seeking approval for the PWR on its Sizewell site based on Westinghouse designs. The very thorough enquiry which lasted for more than a year was completed in March and now awaits the Inspector's report and recommendations, which are not expected before the Autumn at the earliest. The CEGB put forward various weighty reasons for the station amongst other things it argued that building a PWR at Sizewell, even in advance of a need for additional capacity, would save more in fuel bills than it would cost to build and run it. Its net effective cost is negative. The opponents have contested demand projections, all aspects of rector costs, fossil fuel price projections (on which benefit estimates are based), safety arguments and other concerns. Their basic line of argument, as one might expect, is that the reactor is not needed now, even if it were it would be uneconomic, or if economic it would be unsafe, or if safe and economic there are other preferable designs, and in any case proliferation considerations or local impacts should rule a nuclear choice out altogether. All these issues are both debated

Report on NPPCI topics in the United Kingdom - March 1986

International Nuclear Information System (INIS)

Cox, R.J.

1986-01-01

The author reviews the activities on nuclear power plant instrumentation and control topics in the United Kingdom. Since the last meeting of the Working Group in May 1984, the activities of the nuclear power industry in the United Kingdom has been dominated, firstly, by the Public Inquiry into the proposal to site a PWR at Sizewell in Suffolk - the public part of the Inquiry finished about a year ago having lasted eighteen months but the report is not now expected for some months. Secondly, all of the first generation of Advanced Gas Cooled Reactors have run up to power. Thirdly, the Prototype Fast Reactor at Dounreay is now reliably producing its design output of 250 MW(e) following major modifications to its steam generators. The new developments of the NPP simulators, computer control and instrumentation systems as well as reactor protection systems are briefly discussed

Current safety issues related to research reactor operation

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

2000-01-01

The Agency has included activities on research reactor safety in its Programme and Budget (P and B) since its inception in 1957. Since then, these activities have traditionally been oriented to fulfil the Agency's functions and obligations. At the end of the decade of the eighties, the Agency's Research Reactor Safety Programme (RRSP) consisted of a limited number of tasks related to the preparation of safety related publications and the conduct of safety missions to research reactor facilities. It was at the beginning of the nineties when the RRSP was upgraded and expanded as a subprogramme of the Agency's P and B. This subprogramme continued including activities related to the above subjects and started addressing an increasing number of issues related to the current situation of research reactors (in operation and shut down) around the world such as reactor ageing, modifications and decommissioning. The present paper discusses some of the above issues as recognised by various external review or advisory groups (e.g., Peer Review Groups under the Agency's Performance Programme Appraisal System (PPAS) or the standing International Nuclear Safety Advisory Group (INSAG)) and the impact of their recommendations on the preparation and implementation of the part of the Agency's P and B relating to the above subject. (author)

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Helium production in mixed spectrum reactor-irradiated pure elements

International Nuclear Information System (INIS)

Kneff, D.W.; Oliver, B.M.; Skowronski, R.P.

1986-01-01

The objectives of this work are to apply helium accumulation neutron dosimetry to the measurement of neutron fluences and energy spectra in mixed-spectrum fission reactors utilized for fusion materials testing, and to measure helium generation rates of materials in these irradiation environments. Helium generation measurements have been made for several Fe, Cu Ti, Nb, Cr, and Pt samples irradiated in the mixed-spectrum High Flux Isotope Reactor (HFIR) and Oak Ridge Research Reactor (ORR) at the Oak Ridge National Laboratory. The results have been used to integrally test the ENDF/B-V Gas Production File, by comparing the measurements with helium generation predictions made by Argonne National Laboratory using ENDF/B-V cross sections and adjusted reactor spectra. The comparisons indicate consistency between the helium measurements and ENDF/B-V for iron, but cross section discrepancies exist for helium production by fast neutrons in Cu, Ti, Nb, and Cr (the latter for ORR). The Fe, Cu, and Ti work updates and extends previous measurements

Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

Research and development of a super fast reactor (12). Considerations for the reactor characteristics

International Nuclear Information System (INIS)

Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

2008-01-01

A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

Safety of next generation power reactors

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

Safe Management Of Fast Reactors: Towards Sustainability

International Nuclear Information System (INIS)

Dreimanis, Andrejs

2015-01-01

An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

1990-09-01

Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

A review of the UK fast reactor programme, March 1981

International Nuclear Information System (INIS)

Smith, R.D.

1981-01-01

A reduction in electricity sales over the last year had led to some fossil-fuelled stations being prematurely retired and has postponed the start of some new stations. Nevertheless the main programme for the building of 1 5 G We of thermal reactors during the next ten years remains unaltered and as summarised in last year's review. A formal request to build the first PWR at Sizewell in Suffolk has been presented by the CEGB to the government. This is expected to lead to a Public Inquiry within the next 12-18 months. The major contracts for building the AGR stations at Torness and Heysham were placed recently. Reduced projections for electricity demand up to the end of the century have also contributed to a delay in the government's response to the recommendation by the industry that the Commercial Demonstration Fast Reactor (CDFR) should be built to ensure that the option for commercial LMFBR should be demonstrated and maintained. A government statement is now expected before the end of 1981. Fast breeder reactors are expected to be required in the UK electrical supply system by about the turn of the century. Series ordering will be preceded by construction and operation of the CDFR, of a design suitable in all basic features for replication in programme reactors. The National Nuclear Corporation (NNC) has continued the development of a CDFR design having the required operational safety and economic characteristics. The basic design concept is now nearing completion following investigation of a number of alternatives. Some of the more important features of this design, namely the core, primary circuit and reactor cooling systems. steam cycle and boilers, and overall plant and station layout are described in this review. As a result of increased understanding of sodium/water reaction behaviour, development of manufacturing and inspection techniques and experience in plugging and repair of tubes containing leaking welds, coupled with the preference for a once through

A review of the UK fast reactor programme, March 1981

Energy Technology Data Exchange (ETDEWEB)

Smith, R D [Risley Nuclear Power Development Establishment, Risley, Warrington (United Kingdom)

1981-05-01

A reduction in electricity sales over the last year had led to some fossil-fuelled stations being prematurely retired and has postponed the start of some new stations. Nevertheless the main programme for the building of 1 5 G We of thermal reactors during the next ten years remains unaltered and as summarised in last year's review. A formal request to build the first PWR at Sizewell in Suffolk has been presented by the CEGB to the government. This is expected to lead to a Public Inquiry within the next 12-18 months. The major contracts for building the AGR stations at Torness and Heysham were placed recently. Reduced projections for electricity demand up to the end of the century have also contributed to a delay in the government's response to the recommendation by the industry that the Commercial Demonstration Fast Reactor (CDFR) should be built to ensure that the option for commercial LMFBR should be demonstrated and maintained. A government statement is now expected before the end of 1981. Fast breeder reactors are expected to be required in the UK electrical supply system by about the turn of the century. Series ordering will be preceded by construction and operation of the CDFR, of a design suitable in all basic features for replication in programme reactors. The National Nuclear Corporation (NNC) has continued the development of a CDFR design having the required operational safety and economic characteristics. The basic design concept is now nearing completion following investigation of a number of alternatives. Some of the more important features of this design, namely the core, primary circuit and reactor cooling systems. steam cycle and boilers, and overall plant and station layout are described in this review. As a result of increased understanding of sodium/water reaction behaviour, development of manufacturing and inspection techniques and experience in plugging and repair of tubes containing leaking welds, coupled with the preference for a once through

Political-social reactor problems at Berkeley

International Nuclear Information System (INIS)

Little, G.A.

1980-01-01

For better than ten years there was little public notice of the TRIGA reactor at UC-Berkeley. Then: a) A non-student persuaded the Student and Senate to pass a resolution to request Campus Administration to stop operation of the reactor and remove it from campus. b) Presence of the reactor became a campaign-issue in a City Mayoral election. c) Two local residents reported adverse physical reactions before, during, and after a routine tour of the reactor facility. d) The Berkeley City Council began a study of problems associated with radioactive material within the city. e) Friends Of The Earth formally petitioned the NRC to terminate the reactor's license. Campus personnel have expended many man-hours and many pounds of paper in responding to these happenings. Some of the details are of interest, and may be of use to other reactor facilities. (author)

Nuclear propulsion apparatus with alternate reactor segments

International Nuclear Information System (INIS)

Szekely, T.

1979-01-01

Nuclear propulsion apparatus comprising: (a) means for compressing incoming air; (b) nuclear fission reactor means for heating said air; (c) means for expanding a portion of the heated air to drive said compressing means; (d) said nuclear fission reactor means being divided into a plurality of radially extending segments; (e) means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and (f) means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus. 12 claims

Management of research reactor ageing

International Nuclear Information System (INIS)

1995-03-01

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

Management of research reactor ageing

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

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

Emergency cooling of presurized water reactor

International Nuclear Information System (INIS)

Sykora, D.

1981-01-01

The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

In-service inspection of pool type research reactors

International Nuclear Information System (INIS)

Rajamani, K.

2002-01-01

In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

Grouping of HLW in partitioning for B/T (burning and/or transmutation) treatment with neutron reactors based on three criteria

International Nuclear Information System (INIS)

Kitamoto, Mulyanto; Kitamoto, Asashi

1995-01-01

A grouping concept of HLW in partitioning for B/T (burning and/or transmutation) treatment by fission reactor was developed in order to improve the disposal in waste management from the safety aspect. The selecting and grouping concept was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, and trace quantity of Cf, etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remains of HLW), judging from the three criteria for B/T treatment, based on (1) the concept of the potential risk estimated by the hazard index for long-term tendency based on ALI (2) the concept of the relative dose factor related to the adsorbed migration rate transferred through ground water, and (3) the concept of the decay acceleration factor, the burning and/or transmutation characteristics for recycle B/T treatment. (author)

IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Alcala, F.; Di Meglio, A.F.

1995-01-01

This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Gregory, A.R.

1982-11-01

The procedure for decommissioning a CEGB nuclear power station is described. Regulatory and licensing procedures in the UK are first listed. The principal sources of radioactivity in the station after final shutdown are classified. The three stages of the decommissioning procedure are then described. Finally the following topics are dealt with briefly: the management of decommissioning wastes, radiological protection during the operation, possible faults arising having radiological significance, design for decommissioning and costs. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Matthews, R.R.

1982-11-01

Descriptions are given of the CEGB's organizational structure and the policies and practices developed within the Board to achieve safety in its nuclear plants. The division of nuclear safety responsibilities within the CEGB and between the various UK nuclear organizations is outlined. The achievement of nuclear safety and radiological protection in design and in operation is discussed. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Davies, C.H.

1982-11-01

The CEGB's case for constructing new generating plant is considered. Firstly forecasting procedures within the electricity supply industry are outlined. Assessments of electricity demand in the medium term are discussed. These are relevant to a capacity need argument and the economic estimates underlying them. Longer term assessments which are primarily related to an economic need argument are then considered. Work on world economic growth, UK economic growth, UK energy demand and electricity demand on the CEGB is set out. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Townsend, M.

1982-11-01

The organizational arrangements for procuring uranium supplies and the interfaces with the international uranium industry are outlined. Forecasts of WOCA (World Outside Communist Areas) uranium resources and demand are presented together with prices derived from these forecasts. CEGB requirements are reviewed against this backcloth and its existing supply arrangements, activities and long term procurement strategy are discussed. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Bonnell, J.A.

1982-11-01

An account is given of the biological effects of exposure to ionizing radiation and of the sources of radiation burden to the population. A brief history of the ICRP follows. Its latest recommendations are outlined and their endorsement in the UK is described. The work of the occupational health services in the protection of radiation workers and the CEGB's research into health physics and radiobiology are discussed. Finally a comparison of the risks from exposures at the CEGB's occupational design criterion with naturally occuring rates of cancer death and genetic effects is presented. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Flowers, R.

1982-11-01

The role of NIREX (the Nuclear Industry Radioactive Waste Executive) in planning for the disposal of intermediate level waste and low level waste is explained. An inventory of such wastes in the UK is given and future arisings are estimated. Radioactive waste disposal routes currently available in the UK are described. The need for NIREX to increase these disposal facilities and plans for future operations are discussed. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Singleton, L.

1982-11-01

Details are given of the transport of flasks of irradiated fuel by British Rail. Arrangements which have been made for safety on the railways are discussed. Particular arrangements for flask safety and emergency procedures laid down in case of a rail accident are outlined. Extracts from British Railways publications are included as Appendices to the proof. These publications are: the British Railways Rule Book, the Working Manual for Rail Staff, the Board's Byelaws and the List of Dangerous Goods Conditions of Acceptance. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Hughes, P.R.

1982-11-01

Forecasts of the future price and availability of fossil fuels in the UK are presented. Firstly the assumptions and methodology used in constructing three world energy cases for 1980-2030 are outlined. Three world fossil fuel price projections are then derived. The assessment of the prospects for UK fossil fuel supply is discussed and a detailed analysis of the coal industry is made. Fossil fuel prices to the CEGB and final customers in the UK are estimated for each of the five UK economic scenarios. Finally a comparison is made between CEGB projections and other studies. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Womack, G.J.

1982-11-01

The safe transport of irradiated fuel is discussed. Firstly the amount of fuel to be handled, the storage of irradiated fuel at the station, the design of the flask to be used and its cost, and the operations and maintenance involved in operating a flask are considered. The regulations for the safe transport of radioactive material are then described. Possible accident situations and their radiological consequences are discussed. It is concluded that the risk to the population and the environment is very small. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Ward, D.A.

1982-11-01

Descriptions are given of the historical development of the definition and understanding of a LOCA transient, the various types of LOCAs which can be postulated, the faults that have been analysed, the relevant safeguards systems which are designed to limit the consequences of a LOCA and the analytical methods which are used. (U.K.)

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Pepper, R.B.

1982-11-01

The practice of radiological protection within the CEGB is described. Health physics at nuclear power stations is discussed. Radiation exposure of operational staff at a PWR and radiation exposure of members of the public are considered. Finally arrangements for dealing with radiological hazards arising during a site emergency are described. (U.K.)

Material for shutting down gas cooled nuclear reactors

International Nuclear Information System (INIS)

Jackson, F.

1977-01-01

Some disadvantage of conventional emergency shutdown means for nuclear reactors employing a supply of B steel shot or B powder are mentioned. With regard to B powder it is stated that there is some uncertainty as to whether the powder once dispersed into the core will settle in the active part of the core in sufficient quantities to ensure shutdown. The system described aims to avoid these disadvantages. Pellets are provided comprising a solid neutron poison material and a solid organic substance that remains solid at the relatively low temperature normally expected to prevail in the reactor coolant channel away from the reactor core. The organic substance melts at a higher temperature expected to prevail in the coolant channel within the core., and is adherent on melting to the coolant channel wall and to the solid neutron poison, being thus capable of causing adherence of the latter to the coolant channel wall in the reactor core. The pellets are preferably given a moisture resistant coating to prevent them sticking together and to impart free flowing characteristics. The neutron poison may consist of B, Cd, Gd, or their compounds, and for the coating a suitable polymer may be used. Steel filings may be incorporated in the pellets to aid easy flowing under gravity. Examples of manufacture of the pellets are given. (U.K.)

A series of lectures on operational physics of power reactors

International Nuclear Information System (INIS)

Mohanakrishnan, P.; Rastogi, B.P.

1982-01-01

This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

Reactor protection system

International Nuclear Information System (INIS)

Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

1977-10-01

The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

On the structure of Lattice code WIMSD-5B

International Nuclear Information System (INIS)

Kim, Won Young; Min, Byung Joo

2004-03-01

The WIMS-D code is a freely available thermal reactor physics lattice code used widely for thermal research and power reactor calculation. Now the code WIMS-AECL, developed on the basis of WIMS-D, has been used as one of lattice codes for the cell calculation in Canada and also, in 1998, the latest version WIMSD-5B is released for OECD/NEA Data Bank. While WIMS-KAERI was developed and has been used, originated from WIMS-D, in Korea, it was adjusted for the cell calculation of research reactor HANARO and so it has no confirmaty to CANDU reactor. Therefore, the code development applicable to cell calculation of CANDU reactor is necessary not only for technological independence and but also for the establishment of CANDU safety analysis system. A lattice code WIMSD-5B was analyzed in order to set the system of reactor physics computer codes, to be used in the assessment of void reactivity effect. In order to improve and validate WIMSD-5B code, the analysis of the structure of WIMSD-5B lattice code was made and so its structure, algorithm and the subroutines of WIMSD-5B were presented for the cluster type and the pij method modelling the CANDU-6 fuel

Investigation of small and modular-sized fast reactor

International Nuclear Information System (INIS)

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

2000-06-01

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

New ceramics for nuclear industry. Case of fission and fusion reactors

International Nuclear Information System (INIS)

Yvars, M.

1979-10-01

The ceramics used in the nuclear field are described as is their behaviour under radiation. 1) Power reactors - nuclear fission. Ceramics enter into the fabrication of nuclear fuels: oxides, carbides, uranium or plutonium nitrides or oxy-nitrides. Silicon carbide SiC is used for preparing the fuels of helium cooled high temperature reactors. Its use is foreseen in the design of gas high temperature gas thermal exchangers, as is silicon nitride (Si 3 N 4 ). In the materials for safety or control rods, the intense neutron flows induce nuclear reactions which increase the temperature of the neutron absorbing material. Boron carbide B 4 C, rare earth oxides Ln 2 O 3 , or B 4 C-Cu or B 4 C-Al cermets are employed. Burnable poison materials are formed of Al 2 O 3 -B 4 C or Al 2 O 3 -Ln 2 O 3 cermets. The moderators of thermal neutron reactors are in high purety polycrystalline graphite. For the thermal insulation of reactor vessels and jackets, honeycomb ceramics are used as well as ceramic fibres on an increasing scale (kaolin, alumina and other fibres). 2) fusion reactors (Tokomak). These require refractory materials with a low atomic number. Carbon fibres, boron carbide, some borons (Al B 12 ), silicon nitrides and oxy-nitrides and high density alumina are the substances considered [fr

Reactor building

International Nuclear Information System (INIS)

Ebata, Sakae.

1990-01-01

At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

Examination of corrosion on primary selector valve bellows

International Nuclear Information System (INIS)

Rickards, G.K.

1975-07-01

The stainless steel bellows of the primary selector valves from the burst can detection system of the Sizewell 'A' reactor were found to have spots of corrosion. These corrosion spots were thought to be caused by the cleaning process employed during manufacture. Samples subjected to the manufacturing cleaning process were examined in the scanning electron microscope equipped with an X-ray energy dispersive analysis system. The corrosion was shown to be associated with the acid cleaning process employed. Deposits were also left on samples not acid cleaned and it is suggested that these have come from contaminated washing water. (author)

WWER-1000 reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series 12, 'Reactor Simulator Development' (2001). Course material for workshops using a pressurized water reactor (PWR) Simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003) and Training Course Series No. 23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation. N. V. Tikhonov and S. B. Vygovsky of the Moscow Engineering and Physics Institute prepared this report for the IAEA

Pellet bed reactor for nuclear thermal propelled vehicles

International Nuclear Information System (INIS)

El-Genk, M.; Morley, N.J.; Haloulakos, V.E.

1991-01-01

The Pellet Bed Reactor (PeBR) concept is capable of operating at a high power density of up to 3.0 kWt/cu cm and an exit hydrogen gas temperature of 3000 K. The nominal reactor thermal power is 1500 MW and the reactor core is 0.80 m in diameter and 1.3 m high. The nominal PeBR engine generates a thrust of approximately 315 kN at a specific impulse of 1000 s for a mission duration to Mars of 250 days requiring a total firing time of 170 minutes. Because of its low diameter-to-height ratio, PeBR has enough surface area for passive removal of the decay heat from the reactor core. The reactor is equipped with two independent shutdown mechanisms; 8-B4C safety rods and 26 BeO/B4C control drums; each system is capable of operating and scraming the reactor safely. Due to the absence of core internal support structures, the PeBR can be fueled and refueled in orbit using the vacuum of space. These unique features of the PeBR provide for safety during launch, simplicity of handling, deployment, and end-of-life disposal, and vehicle extended lifetime. 11 refs

Chooz-B1, the new Electricite de France PWR: calculation scheme of neutron leakages from the reactor cavity

International Nuclear Information System (INIS)

Champion, G.; Thiriet, A.; Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.

1987-04-01

A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable

Valve testing for UK PWR safety applications

International Nuclear Information System (INIS)

George, P.T.; Bryant, S.

1989-01-01

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

Catalytic Reactor for Inerting of Aircraft Fuel Tanks

Science.gov (United States)

1974-06-01

Aluminum Panels After Triphase Corrosion Test 79 35 Inerting System Flows in Various Flight Modes 82 36 High Flow Reactor Parametric Data 84 37 System...AD/A-000 939 CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS George H. McDonald, et al AiResearch Manufacturing Company Prepared for: Air Force...190th Street 2b. GROUP Torrance, California .. REPORT TITLE CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS . OESCRIP TIVE NOTEs (Thpe of refpoft

Fusion reactor materials

International Nuclear Information System (INIS)

Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

1980-01-01

Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

Evaluation of core compositions for use in breed and burn reactors and limited-separations fuel cycles

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2013-01-01

Highlights: ► Calculated minimum burnup and irradiation damage for B and B reactor compositions. ► Computed doubling time of fuel cycles using B and B reactors and no chemical separations. ► Determined sensitivity of doubling time to using melt refining vs. direct reuse. ► Examined tradeoff between power density and neutronics for different coolants. - Abstract: Previously developed methods for analyzing breed-and-burn (B and B) reactors are applied to a wide range of core compositions. The compositions studied include different fuel types, steel and silicon carbide structure, and sodium, lead/lead bismuth eutectic (LBE), and gas coolants. These compositions are evaluated for use in “minimum burnup” B and B reactors in which it is assumed that blocks comprising the core can be shuffled in all three dimensions to flatten out non-uniformities in burnup. The two figures of merit evaluated are the minimum irradiation damage requirement and reactor fleet doubling time. To minimize irradiation damage, gas coolants perform best, followed by lead/LBE then sodium. High uranium-content metal fuel outperforms compound fuels, and different types of steel are similar and perform slightly better than silicon carbide. Once-through irradiation damage requirements can be surprisingly modest in minimum burnup B and B reactors, with a wide range of compositions viable at irradiation damage levels 50% higher than existing materials data. Doubling times were calculated for a reactor fleet consisting of B and B reactors operating in a limited-separations fuel cycle; i.e., a fuel cycle with no chemical separation of actinides. The effects of different cooling times and removal of fission products using a melt refining process are evaluated. To minimize doubling time, sodium cooled compositions perform best because they are able to achieve core power densities several times larger than compositions using other coolants. A hypothetical sodium-cooled core composition with high

Description of the advanced gas cooled type of reactor (AGR)

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

1996-11-01

The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

Description of the advanced gas cooled type of reactor (AGR)

International Nuclear Information System (INIS)

Nonboel, E.

1996-11-01

The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

Methods in nuclear reactors calculations; Metodos de calculo en reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Velarde, G

1966-07-01

Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P{sub l}; B{sub l}; M{sub l}; S{sub n} and discrete ordinates approximations. (Author)

Reproduction of the RA reactor fuel element fabrication

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr

Neutron noise in nuclear reactors

International Nuclear Information System (INIS)

Blaquiere, A.; Pachowska, R.

1961-06-01

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

iB1350 no.1. A generation III.7 reactor after the Fukushima Daiichi accident

International Nuclear Information System (INIS)

Sato, Takashi; Matsumoto, Keiji; Hosomi, Kenji; Kojima, Yoshihiro; Taguchi, Keisuke

2017-01-01

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. The NSSS is exactly the same as that of the current ABWR. As for safety design, it has a double cylinder RCCV (Mark W containment) and in-depth hybrid safety systems (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, an innovative core catcher (iCC), a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. While the conventional PCCS can never cool the S/P, the iPCCS can automatically cool the S/P directly even in a DBA LOCA. That makes it possible for the iB1350 to optimize the active safety systems for a DBA. Sato came up with several optimized configurations of the IDHS that are expected to achieve further cost reduction and enhance its reliability resulting from passive feature of the iPCCS. The IC/iPCCS pool has enough capacity for 7 day grace period. The IC/iPCCS heat exchangers, the core and the spent fuel pool are

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Research nuclear reactor RA - Annual report 1992

International Nuclear Information System (INIS)

Sotic, O.

1992-12-01

Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

Possible physics modifications to CIRUS reactor core for improved reactor utilization

International Nuclear Information System (INIS)

John, Benjamin; Khosla, S.K.; Narain, Rajendra.

1976-01-01

Two fuelling schemes for uprating the neutron flux in CIRUS reactor at Trombay, are studied. One scheme employs enriched uranium-aluminium alloy boosters, the second envisages employing thorium oxide enriched with 0.2% plutonium oxide. It is seen that the second scheme has the potential of in-situ thorium utilization. (M.G.B.)

How power is generated in a nuclear reactor

International Nuclear Information System (INIS)

Swaminathan, V.

1978-01-01

Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2002-01-01

The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

Reactors. Nuclear propulsion ships

International Nuclear Information System (INIS)

Fribourg, Ch.

2001-01-01

This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

International Nuclear Information System (INIS)

Huda, M.Q.

2006-01-01

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

Set of rules SOR 2 licensing of nuclear reactors

International Nuclear Information System (INIS)

1976-05-01

This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

Accident dynamics of LR-0 reactor

International Nuclear Information System (INIS)

Vorisek, M.; Tinka, I.

1981-01-01

The results are given of calculating the accident dynamics of the LR-0 light water experimental zero power reactor. Calculations of the time dependence of power, the total released energy, the temperature of fuel and its cladding were made using program FATRAP for different values of the total inserted reactivity. Using the results, an analysis is made of hypothetic accident states of the LR-0 reactor. The results are shown graphically. (J.B.)

Environmentally assisted cracking in light water reactors

International Nuclear Information System (INIS)

Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Soppet, W.K.; Shack, W.J.

1992-03-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

Dynamic problems of power reactors and analogic devices

International Nuclear Information System (INIS)

Braffort, P.

1955-01-01

The raise of the nuclear physics came with heavy mathematical developments. The analogical installations became especially useful for precise calculations of parameters which depend the running of a reactor. They permit between other to study of kinetic problems and especially ''cybernetics'' of nuclear reactors. It doesn't make a doubt that their use will become widespread, not only in the calculations laboratories, in services for servo-mechanisms study, but also in the control panels of the reactors themselves. (M.B.) [fr

Bioaccumulation and toxicity assessment of irrigation water contaminated with boron (B) using duckweed (Lemna gibba L.) in a batch reactor system.

Science.gov (United States)

Türker, Onur Can; Yakar, Anıl; Gür, Nurcan

2017-02-15

The present study assesses ability of Lemna gibba L. using a batch reactor approach to bioaccumulation boron (B) from irrigation waters which were collected from a stream in largest borax reserve all over the world. The important note that bioaccumulation of B from irrigation water was first analyzed for first time in a risk assessment study using a Lemna species exposed to various B concentrations. Boron toxicity was evaluated through plant growth and biomass production during phytoremediation process. The result from the present experiment indicated that L. gibba was capable of removing 19-63% B from irrigation water depending upon contaminated level or initial concentration. We also found that B was removed from aqueous solution following pseudo second order kinetic model and Langmuir isotherm model better fitted equilibrium obtained for B phytoremediation. Maximum B accumulation in L. gibba was determined as 2088mgkg -1 at average inflow B concentration 17.39mgL -1 at the end of the experiment. Conversely, maximum bioconcentration factor obtained at lowest inflow B concentrations were 232 for L. gibba. The present study suggested that L. gibba was very useful B accumulator, and thus L. gibba-based techniques could be a reasonable phytoremediation option to remove B directly from water sources contaminated with B. Copyright © 2016 Elsevier B.V. All rights reserved.

Calculation of the source term for a S1B-sequence at a VVER-1000 type reactor. Part 1

International Nuclear Information System (INIS)

Sdouz, G.

1990-10-01

The behaviour of the source term in a VVER-1000 type reactor is calculated using the 'Source Term Code Package' (STCP). The input data are based on the russian plant Zaporozhye-5. The selected accident sequence is a small break LOCA in the hot leg followed by loss offsite and onsite electric power (S 1 B-sequence). According to the course of the calculation the results are presented and analyzed for each program. Except for the noble gases all release fractions are lower than 10 -4 . 18 refs., 10 tabs. (Author)

Reactor modification, preparation and operation

International Nuclear Information System (INIS)

Weill, J.; Furet, J.; Baillet, J.; Donvez, G.; Duchene, J.; Gras, R.; Mercier, R.; Chenouard, J.; Leconte, J.

1962-01-01

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system

Reactor modification, preparation and operation

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

1962-03-01

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

Reactor modification, preparation and operation

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

1962-03-15

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

Evaluation of nuclear reactor based activation analysis techniques

International Nuclear Information System (INIS)

Obrusnik, I.; Kucera, J.

1977-09-01

A survey is presented of the basic types of activation analysis applied in environmental control. Reactor neutron activation analysis is described (including the reactor as a neutron source, sample activation in the reactor, methodology of neutron activation analysis, sample transport into the reactor and sample packaging after irradiation, instrumental activation analysis with radiochemical separation, data measurement and evaluation, sampling and sample preparation). Sources of environmental contamination with trace elements, sampling and sample analysis by neutron activation are described. The analysis is described of soils, waters and biological materials. Methods are shown of evaluating neutron activation analysis results and of their interpretation for purposes of environmental control. (J.B.)

Coal price prospects and availability of coal in the U.K. power generation market

International Nuclear Information System (INIS)

Parker, M.J.

1983-02-01

The availability and cost of National Coal Board coal is discussed with respect to the CEGB's economic case for Sizewell B nuclear power station. It is concluded that an investment which depended for its viability on an early or rapid escalation in international coal prices, or upon this escalation continuing indefinitely into the future, would not be sound. (U.K.)

Steam explosions in light water reactors

International Nuclear Information System (INIS)

1981-01-01

The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

1997-01-01

In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

International Nuclear Information System (INIS)

Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

1993-01-01

The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

Energy Technology Data Exchange (ETDEWEB)

Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

2016-09-15

Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

Iiyoshi, A.; Momota, H.; Motojima, O.

1994-01-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author).

Reactor protection system. Revision 1

International Nuclear Information System (INIS)

Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

International Nuclear Information System (INIS)

Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

2016-01-01

Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

Science.gov (United States)

You, Yan; Yoshida, Katsumi; Yano, Toyohiko

2018-05-01

Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

Massive use of pulverized fuel ash in concrete for the construction of a U.K. power station

International Nuclear Information System (INIS)

Davies, D.R.; Kitchener, J.N.

1996-01-01

This paper describes the incorporation and benefits of pulverized fuel ash (PFA) in nearly 620,000 m 2 of concrete used in the construction of the UK's first commercial pressurized light water nuclear reactor power station, Sizewell B, Suffolk. Overall nearly 100,000 t of PFA, 1,300 t of sintered fly ash lightweight aggregate, and nearly 137,000 t of Ordinary Portland Cement (OPC) were used in the works. Generally in the construction of the main power station buildings, structural concrete with a characteristic strength of 45 N/mm 2 was placed and included fly ash as 40% of the cementitious component. Also, concrete with fly ash as 50% of the cementitious component was placed as mass fill. The program of research is reported, including sulfate resistance, heat of hydration,elastic properties, alkali-silica reaction, and long-term strength test results. Test and field results are presented and discussed. Strict quality assurance procedures were enforced and statistical summaries are offered to give an insight into the quality of control exercised. Test work and long-term results are also outlined

A user friendly database for use in ALARA job dose assessment

Energy Technology Data Exchange (ETDEWEB)

Zodiates, A.M.; Willcock, A. [Cheshire, England (United Kingdom)

1995-03-01

The pressurized water reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant (SNUPPS). This design was developed to meet the United Kingdom requirements and these improvements are embodied in the Sizewell B plant which will start commercial operation in 1994. A user-friendly database was developed to assist the station in the dose and ALARP assessments of the work expected to be carried out during station operation and outage. The database stores the information in an easily accessible form and enables updating, editing, retrieval, and searches of the information. The database contains job-related information such as job locations, number of workers required, job times, and the expected plant doserates. It also contains the means to flag job requirements such as requirements for temporary shielding, flushing, scaffolding, etc. Typical uses of the database are envisaged to be in the prediction of occupational doses, the identification of high collective and individual dose jobs, use in ALARP assessments, setting of dose targets, monitoring of dose control performance, and others.

Public and parliamentary examination of controversial projects. Address to the Parliamentary Group for Energy Studies, 25th January 1989

Energy Technology Data Exchange (ETDEWEB)

Kemp, R.; O' Riordan, T.; Purdue, M.

1989-04-01

In this paper we consider the underlying significance of the way in which major energy projects are subjected to public scrutiny in the United Kingdom. In particular, we will focus on reforms in the procedures at big public inquiries which have been introduced following the Sizewell B Pressurised Water Reactor Inquiry, and on the efficacy of alternative, Parliamentary-based procedures for dealing with major projects. Setting these thoughts into greater relief, we turn to the identifiable problems associated with both radioactive waste disposal and more briefly the River Severn Tidal Barrage proposals. We outline some possible ways in which the public examination of these proposals might better proceed. In general, we argue that there is a great need to co-ordinate the roles of Parliamentary Select Committees and public inquiries in order to provide a more thorough-going and ultimately publicly acceptable scrutiny of the major energy projects. A balance needs to be struck between the need for fairness on the one hand and efficient use of public and private funds on the other. (author).

Neutronic study of the two french heavy water reactors

International Nuclear Information System (INIS)

Horowitz, J.

1955-01-01

The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

Benchmark testing of Canadol-2.1 for heavy water reactor

International Nuclear Information System (INIS)

Liu Ping

1999-01-01

The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

1994-01-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Radiation detectors for reactors

International Nuclear Information System (INIS)

Balagi, V.

2005-01-01

Detection and measurement of radiation plays a vital role in nuclear reactors from the point of view of control and safety, personnel protection and process control applications. Various types of radiation are measured over a wide range of intensity. Consequently a variety of detectors find use in nuclear reactors. Some of these devices have been developed in Electronics Division. They include gas-filled detectors such as 10 B-lined proportional counters and chambers, fission detectors and BF 3 counters are used for the measurement of neutron flux both for reactor control and safety, process control as well as health physics instrumentation. In-core neutron flux instrumentation employs the use detectors such as miniature fission detectors and self-powered detectors. In this development effort, several indigenous materials, technologies and innovations have been employed to suit the specific requirement of nuclear reactor applications. This has particular significance in view of the fact that several new types of reactors such as P-4, PWR and AHWR critical facilities, FBTR, PFBR as well as the refurbishment of old units like CIRUS are being developed. The development work has sought to overcome some difficulties associated with the non-availability of isotopically enriched neutron-sensing materials, achieving all-welded construction etc. The present paper describes some of these innovations and performance results. (author)

Paracantor: A two group, two region reactor code

Energy Technology Data Exchange (ETDEWEB)

Stone, Stuart

1956-07-01

Paracantor I a two energy group, two region, time independent reactor code, which obtains a closed solution for a critical reactor assembly. The code deals with cylindrical reactors of finite length and with a radial reflector of finite thickness. It is programmed for the 1.B.M: Magnetic Drum Data-Processing Machine, Type 650. The limited memory space available does not permit a flux solution to be included in the basic Paracantor code. A supplementary code, Paracantor 11, has been programmed which computes fluxes, .including adjoint fluxes, from the .output of Paracamtor I.

Collective study of plans and feature of the reactor for medical usage

International Nuclear Information System (INIS)

1980-01-01

In order to construct the reactor for medical usage comparative studies of irradiating apparatus were performed, and plans to construct medical reactors were constructed by 20 groups consisted of universities, institutes, and companies. As for facilities, a research for TRIGA type reactor, combination of a reactor and an accelerator, and problems in constructing a reactor were investigated. Examinations, with regard to flux, were carried out from the view point of flux variation due to absorber and monitoring thermal neutron dose, while irradiating boron. Some physical problems of neutron detector, neutron source, and preparing enriched isotopes of 10 B were also studied. Analysis of boron was developed by utilization of α autoradiography, synthesis of Na 2 10 B 12 H 11 SH, and enrichment of 10 B. In the field of biomedical science, application of neutron capture method to cerebral tumors, histo-immunological study of the normal brain by enzyme antibody method, and selective radiotherapy of malignant skin tumors were examined using animals. Radiotherapy by neutron capture was carried out to the patients with various tumors, and the remote anesthetization was also tried. (Nakanishi, T.)

Lower activation materials and magnetic fusion reactors

International Nuclear Information System (INIS)

Conn, R.W.; Bloom, E.E.; Davis, J.W.; Gold, R.E.; Little, R.; Schultz, K.R.; Smith, D.L.; Wiffen, F.W.

1984-01-01

Radioactivity in fusion reactors can be effectively controlled by materials selection. The detailed relationship between the use of a material for construction of a magnetic fusion reactor and the material's characteristics important to waste disposal, safety, and system maintainability has been studied. The quantitative levels of radioactivation are presented for many materials and alloys, including the role of impurities, and for various design alternatives. A major outcome has been the development of quantitative definitions to characterize materials based on their radioactivation properties. Another key result is a four-level classification scheme to categorize fusion reactors based on quantitative criteria for waste management, system maintenance, and safety. A recommended minimum goal for fusion reactor development is a reference reactor that (a) meets the requirements for Class C shallow land burial of waste materials, (b) permits limited hands-on maintenance outside the magnet's shield within 2 days of a shutdown, and (c) meets all requirements for engineered safety. The achievement of a fusion reactor with at least the characteristics of the reference reactor is a realistic goal. Therefore, in making design choices or in developing particular materials or alloys for fusion reactor applications, consideration must be given to both the activation characteristics of a material and its engineering practicality for a given application

Cobalt-60 control in Ontario Hydro reactors

International Nuclear Information System (INIS)

Lacy, C.S.

1988-01-01

This paper discusses the impact of specifying reduced Cobalt-59 in the primary heat transport circuit materials of construction on the radiation fields developed around the primary circuit. An eight-fold reduction in steam generator radiation fields due to Cobalt-60 has been observed for two identical sets of reactors, one with and one without Cobalt-59 control. The comparison is between eight reactors at the Pickering Nuclear Generating Station (PNGS). Units 5 to 8 (PNGS-B) are identical to Units 1 to 4 (PNGS-A) except that PNGS-B has reduced impurity Cobalt-59 in the alloys of construction and a reduced use of stellite. The effects of chemistry control are also discussed

Environmentally assisted cracking in Light Water Reactors

International Nuclear Information System (INIS)

Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

1993-09-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

SCORPIO: a framework for core surveillance systems

International Nuclear Information System (INIS)

Berg, Oe.; Tsuiki, Makoto

1999-01-01

The first version of the core surveillance system SCORPIO was installed at Unit 2, Ringhals, in 1984. It was implemented on Norsk Data mini-computers with a fully graphical user-interface. The main purpose was to provide a practical tool for reactor operators and reactor physicists for on-line monitoring and predictive analysis of core behaviour. A second version of SCORPIO was developed in 1993-1995 and implemented on Unix workstations. In addition to upgrading the system at Ringhals, the system was installed by Duke Power, USA, on 7 reactors. SCORPIO was also installed on the Sizewell B reactor. Recently a new framework has been developed which further enhances the flexibility and capabilities for implementing core surveillance systems in different types of nuclear power plants. Modules can be added and replaced in an easy manner. It allows fast and reliable communication of data between modules based on the Software Bus tool developed by IFE. Further, the Picasso-3 user interface management system supports efficient implementation of different user interfaces. Both Unix and Windows NT platforms are supported. The new framework has been applied in development and installation of a SCORPIO-VVER version for the Dukovany NPP, Czech Republic. Here it was of particular importance to provide a flexible system for integration of modules originating from different companies. Development of a BWR version is now in progress. This means that SCORPIO will be available for all the major reactor types, and synergy is obtained by application of a common framework both with respect to system implementation and maintenance. By using the SCORPIO framework, the development time is reduced and the maintenance work is carried out more efficiently, compared to developing systems with lower-level tools. For instance, the MMI can be developed and tested independently of the physics modules

Use of the VR-1 ''Vrabec'' training reactor

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

Catalytic Reactor For Oxidizing Mercury Vapor

Science.gov (United States)

Helfritch, Dennis J.

1998-07-28

A catalytic reactor (10) for oxidizing elemental mercury contained in flue gas is provided. The catalyst reactor (10) comprises within a flue gas conduit a perforated corona discharge plate (30a, b) having a plurality of through openings (33) and a plurality of projecting corona discharge electrodes (31); a perforated electrode plate (40a, b, c) having a plurality of through openings (43) axially aligned with the through openings (33) of the perforated corona discharge plate (30a, b) displaced from and opposing the tips of the corona discharge electrodes (31); and a catalyst member (60a, b, c, d) overlaying that face of the perforated electrode plate (40a, b, c) opposing the tips of the corona discharge electrodes (31). A uniformly distributed corona discharge plasma (1000) is intermittently generated between the plurality of corona discharge electrode tips (31) and the catalyst member (60a, b, c, d) when a stream of flue gas is passed through the conduit. During those periods when corona discharge (1000) is not being generated, the catalyst molecules of the catalyst member (60a, b, c, d) adsorb mercury vapor contained in the passing flue gas. During those periods when corona discharge (1000) is being generated, ions and active radicals contained in the generated corona discharge plasma (1000) desorb the mercury from the catalyst molecules of the catalyst member (60a, b, c, d), oxidizing the mercury in virtually simultaneous manner. The desorption process regenerates and activates the catalyst member molecules.

[International Thermonuclear Experimental Reactor support

International Nuclear Information System (INIS)

Dean, S.O.

1990-01-01

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

Pressurized heavy-water reactor safety

International Nuclear Information System (INIS)

Pease, L.; Wilson, R.

1977-09-01

CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

Multigroup perturbation model for kinetic analysis of nuclear reactors

International Nuclear Information System (INIS)

Souza, G.M.

1989-01-01

The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)

Status of spent fuel management in the United Kingdom

International Nuclear Information System (INIS)

Dodds, R.

1996-01-01

Nuclear generating capacity in the UK is static with no units currently under construction following the completion of the Sizewell B PWR. The Government's reviews of nuclear energy policy and radioactive waste management policy have been published following a public consultation procedure, largely with an endorsement of current policies. Nuclear Electric plc (NE) and Scottish Nuclear Limited's (SNL) AGR and PWR stations are to be privatised as two subsidiaries of a holding company, and it is planned that the Magnox stations and their liabilities will be kept in the public sector, initially in a stand alone company but ultimately integrated with BNFL. Prompt reprocessing of all Magnox fuel will continue. NE and SNL have signed contracts for extensive reprocessing of AGR fuel. In addition, SNL has agreed contractual arrangements with BNFL for long term storage of its remaining overlife arisings of AGR fuel and has therefore on commercial ground opted not to dry store their fuel at the reactor site. NE have not yet made a decision on the fate of their AGR fuel not covered by existing reprocessing contracts. No option selection has taken place for PWR fuel. Following the closure of the Dounreay PFR and the withdrawal from the EFR project, the option of recycle of plutonium in fast reactors has been suspended. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

Comparison of Pickering NGS performance with world power reactors, 1977

International Nuclear Information System (INIS)

Buhay, S.

Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

Functional approaches: a new view of nuclear reactors management

International Nuclear Information System (INIS)

Papin, B.

2000-01-01

Since many years a research program on the future reactors command, has been decided by the CEA, in collaboration with EDF and Framatome. This program aims to enhance the reactor safety by a better control of the installation in any exploitation situations. The paper presents the state of the art and the first reflexions. (A.L.B.)

Fatigue crack growth characteristics of a533 brade b glass i plate in an environment of high-temperature primary grade nuclear reactor water

International Nuclear Information System (INIS)

Mager, T.R.; Moon, D.M.; Landes, J.D.

1976-01-01

To characterize the effect of environment on crack growth rate properties of reactor pressure vessel materials, a program was initiated as part of the Heavy Section Steel Technology Program (HSST) to evaluate the effect of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) primary grade coolant environments. The experiments included such variables as frequency, temperature and R ratio. This paper describes the investigation and presents the results of a fracture mechanics evaluation of the fatigue crack growth rate tests of A533 Grade B Class 1 steel plate material in an environment of primary reactor grade water at 550 0 F (288 0 C). A compliance crack growth monitoring technique was utilized to measure the crack growth. The compliance crack length monitor uses a linear variable differential transformer (LVDT) to measure the specimen front face displacement which is converted to crack length by the appropriate compliance calibration curve. The crack growth rate tests were conducted on constant load universal fatigue machines, under sinusoidal tension to tension loading conditions. Tests showed an increase in growth rates at a frequency of 1 cpm over previous results obtained at frequencies of 60 cpm and higher. This increase, the general character of the crack growth rate versus the $DELTA$K curve, and the results from fractographic studies, all indicated that stress corrosion cracking might have occurred for this material and environment. However, a specimen loaded statically in a PWR environment showed no static load crack growth. 13 refs

Blankets for fusion reactors : materials and neutronics

International Nuclear Information System (INIS)

Carvalho, S.H. de.

1980-03-01

The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

Neutron noise measurement technique in a coupled reactor

International Nuclear Information System (INIS)

Genoud, J.P.

1976-01-01

Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

Sampling and Analysis Plan for Disposition of the Standing Legacy Wastes in the 105-B, -D, -H, -KE, and -KW Reactor Buildings

International Nuclear Information System (INIS)

McGuire, J. J.

1999-01-01

This sampling and analysis plan (SAP) presents the rationale and strategy for the sampling and analysis activities that support disposition of legacy waste in the Hanford Site's 105-B, 105-D, 105-H,105-KE, 105-KW Reactor buildings. For the purpose of this SAP, legacy waste is identified as any item present in a facility that is not permanently attached to the facility and is easily removed without the aid of equipment larger than a standard forklift

Sampling and Analysis Plan for Disposition of the Standing Legacy Wastes in the 105-B, -D, -H, -KE, and -KW Reactor Buildings

International Nuclear Information System (INIS)

McGuire, J.J.

1999-01-01

This sampling and analysis plan (SAP) presents the rationale and strategy for the sampling and analysis activities that support disposition of legacy waste in the Hanford Site's 105-B, 105-D, 105-H, 105-KE, 105-KW Reactor buildings. For the purpose of this SAP, legacy waste is identified as any item present in a facility that is not permanently attached to the facility and is easily removed without the aid of equipment larger than a standard forklift

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Science.gov (United States)

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

Methods in nuclear reactors calculations

International Nuclear Information System (INIS)

Velarde, G.

1966-01-01

Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P l ; B l ; M l ; S n and discrete ordinates approximations. (Author)

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

Reactor group constants and benchmark test

Energy Technology Data Exchange (ETDEWEB)

Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

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

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Steam explosions in sodium cooled breeder reactors

International Nuclear Information System (INIS)

Lundell, B.

1982-01-01

Steam explosion is considered a physical process which transport heat from molten fuel to liquid coolant so fast that the coolant starts boiling in an explosion-like manner. The arising pressure waves transform part of the thermal energy to mechanical energy. This can stress the reactor tank and threaten its hightness. The course of the explosion has not been theoretical explained. Experimental results indicate that the probability of steam explosions in a breeder reactor is small. The efficiency of the transformation of the heat of fusion into mechanical energy in substantially lower than the theoretical maximum value. The mechanical stress from the steam explosion on the reactor tank does not seem to jeopardize its tightness. (G.B.)

Recommendations for a restart of Molten Salt Reactor development

International Nuclear Information System (INIS)

Moir, R. W.

2007-01-01

The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 2 32U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-11-01

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

Mechanical core coupling and reactors stability

International Nuclear Information System (INIS)

Suarez Antola, R.

2006-01-01

Structural parts of nuclear reactors are complex mechanical systems, able to vibrate with a set of proper frequencies when suitably excited. Cyclical variations in the strain state of the materials, including density perturbations, are produced. This periodic changes may affect reactor reactivity. But a variation in reactivity affects reactor thermal power, thus modifying the temperature field of the abovementiones materials. If the variation in temperature fields is fast enough, thermal-mechanical coupling may produce fast variations in strain states, and this, at its turn, modifies the reactivity, and so on. This coupling between mechanical vibrations of the structure and the materials of the core, with power oscillations of the reactor, not only may not be excluded a priori, but it seems that it has been present in some stage of the incidents or accidents that happened during the development of nuclear reactor technology. The purpose of the present communication is: (a) To review and generalize some mathematical models that were proposed in order to describe thermal-mechanical coupling in nuclear reactors. (b) To discuss some conditions in which significant instabilities could arise, including large amplitude power oscillations coupled with mechanical vibrations whose amplitudes are too small to be excluded by conventional criteria of mechanical design. Enough Certain aspects of thr physical safety of nuclear power reactors, that are objected by people that opposes to the renaissance of nucleoelectric generation, are discussed in the framework of the mathematical model proposed in this paper [es

The accident of Chernobylsk-4 reactor and its consequences

International Nuclear Information System (INIS)

1986-01-01

This report deals with the particulars of the accident as communicated by the Soviet delegation at an IAEA meeting by the and of August 1986. It was stated that the consequences emanated from the inherent instability of the design of the reactor, the deviation from the safety rules by the operators and the lack of a sight reactor containment. (G.B.)

THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

International Nuclear Information System (INIS)

Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

1988-04-01

THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

Temperature monitoring of gas-cooled reactors

International Nuclear Information System (INIS)

Kaiser, G.E.

1977-01-01

The present paper deals with questions like : a) Why temperature monitoring in high-temperature reactors at all. b) How are the measuring positions arranged and how are the measurements designed. c) What technique of temperature measurement is applied. (RW) [de

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

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

1987-01-01

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

Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

A pellet-clad interaction failure criterion

International Nuclear Information System (INIS)

Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

1983-01-01

A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

Design study of superconducting inductive energy storages for tokamak fusion reactor

International Nuclear Information System (INIS)

1977-08-01

Design of the superconducting inductive energy storages (SC-IES) has been studied. One SC-IES is for the power supply system in a experimental tokamak fusion reactor, and the other in a future practical reactor. Study started with definition of the requirements of SC-IES, followed by optimization of the coil shape and determination of major parameters. Then, the coil and the vessel were designed, including the following: for SC-IES of the experimental reactor, stored energy 10 GJ, B max 8 T, conductor NbTi and size 18 m diameter x 10 m height; for SC-IES of the practical reactor, stored energy 56 GJ, B max 10.5 T, conductor Nb 3 Sn and size 26 m diameter x 15 m height. Design of the coil protection system and an outline of the auxiliary systems (for refrigeration and evacuation) are also given, and further, problems and usefullness of SC-IES. (auth.)

Nuclear data evaluation and group constant generation for reactor analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Lee, Jong Tae; Min, Byung Joo; Gil, Choong Sup [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

1991-01-01

In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi- group constant library using the newly compiled data files and the code systems. As the results of this project, ENDF/B-VI Supplementary File including important nuclides, JENDL-3.1 and JEF-1 were compiled, and ENDF-6 international computer file format for evaluated nuclear data and its processing system NJOY89.31 were tested with ENDF/B-VI data. In order to test an applicability of the newly released data to thermal reactor problems, a number of benchmark calculations were performed, and the results were analyzed. Since preliminary benchmark testing of thermal reactor problems have been made the newly compiled data are expected to be positively used to develop advanced reactors. (Author).

Nuclear data evaluation and group constant generation for reactor analysis

International Nuclear Information System (INIS)

Kim, Jung Do; Lee, Jong Tae; Min, Byung Joo; Gil, Choong Sup

1991-01-01

In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi- group constant library using the newly compiled data files and the code systems. As the results of this project, ENDF/B-VI Supplementary File including important nuclides, JENDL-3.1 and JEF-1 were compiled, and ENDF-6 international computer file format for evaluated nuclear data and its processing system NJOY89.31 were tested with ENDF/B-VI data. In order to test an applicability of the newly released data to thermal reactor problems, a number of benchmark calculations were performed, and the results were analyzed. Since preliminary benchmark testing of thermal reactor problems have been made the newly compiled data are expected to be positively used to develop advanced reactors. (Author)

Probabilistic safety assessment applications and insights

International Nuclear Information System (INIS)

Hitchler, M.J.; Burns, N.L.; Liparulo, N.J.; Mink, F.J.

1987-01-01

The insights gained through a comparison of seven PRA studies (Italian PUN, Sizewell B, Ringhals 2, Millstone 3, Zion 1 and 2, Oconee 3, and Seabrook) included insights regarding the adequacy of the PRA technology utilized in the studies and the potential areas for improvement and insights regarding the adequacy of plant designs and how PRA has been utilized to enhance the design and operation of nuclear power plants. (orig.)

A regulatory perspective on human factors in nuclear power

International Nuclear Information System (INIS)

Whitfield, D.

1987-01-01

This paper sets out the approaches being taken by the United Kingdom Nuclear Installations Inspectorate (NII) to monitoring the application of human factors principles and practice in the UK industry. The role of NII is outlined, the development of human factors concerns is reviewed, the assessment of the Sizewell 'B' safety case is presented as a particular example, and pertinent future developments in the human factors discipline are proposed. (author)

Light-water-reactor coupled neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

Diamond, D.J.

1982-01-01

An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

International Nuclear Information System (INIS)

Kotake, Syoji; Nishikawa, Akira

2005-01-01

Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

Application of NDT and ISI to research reactor in the Czech Republic

International Nuclear Information System (INIS)

Peterka, F.

2001-01-01

Full text: The objectives of the proposed research project are: (a) to review the present status of ISI and NDT to VR-1 and LVR-15 research reactors. (b) to be involved in the development of the ISI programme for VR-1 and LVR-15 research reactors and medium and high power research reactors of WWER type. Dr. Peterka briefly described the activities on the VR-1 and LWR-15 reactors and presented an example of a procedure to apply liquid penetrant testing. (author)

Babcock and Wilcox model for predicting in-reactor densification

International Nuclear Information System (INIS)

Buescher, B.J.; Pegram, J.W.

1977-07-01

The B and W densification model is based on a correlation between in-reactor densification and a thermal resintering test. The densification model has been found to predict in-reactor densification with a remarkable degree of accuracy for fuel pellets operated at heat rates above 5 kW/ft and with considerable conservatism for pellelts operating at heat rates below 5 kW/ft

The influence of the operating schedule of the Greek Research Reactor on the radiological consequences of the reactor

International Nuclear Information System (INIS)

Kollas, J.G.

1986-04-01

The sensitivity of the radiological consequences of the Greek Research Reactor to the operating schedule of the reactor is assessed in this report. The consequences are due to the occurrence of a postulated accident, a 20% core melt loss of coolant accident. Three different operating schedules are considered: (a) the present 8 hrs/day, 5 days/wk schedule, (b) a 16 hrs/day, 5 days/wk schedule, and (c) a continuous operation schedule. The results of the analysis indicate that there is a direct relation between consequences and duration of operation. (author)

Reactor alarm system development and application issues

Energy Technology Data Exchange (ETDEWEB)

Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

1997-09-01

The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

Reactor alarm system development and application issues

International Nuclear Information System (INIS)

Drexler, J.E.; Oicese, G.O.

1997-01-01

The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

Nuclear data and multigroup methods in fast reactor calculations

International Nuclear Information System (INIS)

Gur, Y.

1975-03-01

The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)