Hinkley Point 'C' power station public inquiry: proof of evidence on CEGB policy with privatisation in prospect

International Nuclear Information System (INIS)

Davis, D.A.

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. Part of the evidence submitted to the Inquiry is set out in this report. It concerns the implications for the CEGB's initial case of subsequent government proposals for electricity supply privatization. That case rested on the need for diversity of primary fuel source. Present Government policy endorses the CEGB policy in this regard in the intention to make legislative provision for a quota of electricity generation from non-fossil-fueled sources in a privatized supply industry. (UK)

Hinkley Point 'C' power station public inquiry: proof of evidence on plant parameters

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 overall economics of a nuclear power station depends on many factors which are determined by the design; the effectiveness with which the station is constructed; and the performance of the plant. In this respect the most significant factors are: construction time; capital cost; availability of the plant to produce electricity, taking account of those outages due to either planned or unplanned shutdowns; net electrical power output; and the working life of the plant. In this evidence to the Inquiry, the basis for the values chosen as ''targets'' for these parameters in the design of the plant and the control of the project is set out. The adjustment of the parameters to make them suitable for economic appraisal is explained. The design and project management arrangements are described. (author)

Hinkley Point C: French technology and know-how are welcome in Great Britain

International Nuclear Information System (INIS)

Le Ngoc, B.

2016-01-01

The 2 reactors of the Hinkley Point-C will be the fifth and the sixth EPR units built in the world and will benefit from series production. EDF and its Chinese partner CGN and the British government signed the definitive agreement in september 2016. The project involves an innovative financing: the contract for difference, this contract was negotiated in 2013 and agreed by the European Union in 2014. This contract sets a price for selling electricity: 92.50 pounds (2012) and if the market price is below this price, the British government will have to pay the difference and if the market price is above, EDF/CGN will have to pay the difference to the British Government. The set price includes all the costs from the construction to the management of nuclear waste and via its daily plant operating. The commissioning of the first unit is expected in 2025. The supply chain is today in good working order. The total cost of construction is estimated to be 22 billions euros, 60% of the enterprises working on the construction site will be british and the Hinkley Point C project is expected to generate 7000 new jobs in France. (A.C.)

Measurements of the fuel temperature coefficient of reactivity at Hinkley Point 'B': 1981

International Nuclear Information System (INIS)

George, T.A.

1982-03-01

Measurements of the fuel temperature coefficient of reactivity made at Hinkley Point 'B' AGR in 1981 are described. These measurements follow earlier tests reported in e.g. RD/B/N4846 and are part of a series of measurements designed to support theoretical estimates of the change of fuel temperature coefficient as a function of core irradiation. Low and high power measurements were made at a mean core irradiation of 1170GWD. As previously, the measurements at both power levels show agreement with theoretical predictions to within the estimated experimental errors. Recent measurements (mean core irradiation >500GWD) show evidence of a small systematic difference between measured and theoretical values with the experimental values being approximately equal to 0.1mN/ 0 C more positive than the theoretical ones. The measured value of αsub(U) at high power was -0.64+-0.10mN/ 0 C and the low power value, corrected theoretically to normal operating conditions, was also -0.64+-0.10mN/ 0 C. (author)

Development, testing and installation of prestressing of the PCPV's at Hinkley Point B and Hunterston B

International Nuclear Information System (INIS)

Taylor, S.J.; Eadie, D.McD.

1976-01-01

In order to keep the walls of the PCPVs at Hinkley Point B and Hunterston B as thin as possible it was desirable to have a prestressing system which permitted close spacing of the tendons and anchorages. A seven strand tendon system was adopted in which the strands are individually anchored by the well-established wedge and barrel system. The seven anchorages bear on to a square plate which transmits the anchorage load to the concrete via a cast steel trumpet. These seven strand anchorages are arranged in groups of four and precast into a steel box as an anchorage unit with an ultimate tendon load of about 1050 tonnes. The development leading up to the adoption of this prestressing system and anchorage arrangement is described. Full size proving tests were carried out on the four tendon anchorage box unit, together with an alternative arrangement using a square helix of bonded reinforcement. Information is given on the friction tests carried out on the completed vessel, prestressing of the PCPVs and the corrosion prevention methods adopted. (author)

The CNEN, Flamanville 3 and the English project of Hinkley Point

International Nuclear Information System (INIS)

Reber, Laurent

2014-01-01

The CNEN is the National Centre of Nuclear Equipment. It's an engineering unit of EDF. After a brief presentation of this department which is one of the six engineering departments of EDF, the CNEN manager presents the Flamanville EPR, its design characteristics and objectives (notably in terms of safety and availability). He comments the status of the construction, gives some explanations to the difficulties faced during this construction. He outlines the greater importance given to safety with respect to reactors of previous generations. Then, he addresses the British Hinkley Point project, its context, the requirements expressed by the British nuclear safety authority. As studies must be performed in the host country, the British nuclear engineering organisation is indicated. The content of the contract between EDF and the British is summarized

GAO report and EDF cost revisions reignite debate on Hinkley Point C

Energy Technology Data Exchange (ETDEWEB)

Dalton, David [NucNet, The Independent Global Nuclear News Agency, Brussels (Belgium)

2017-10-15

The announcement by French state-controlled utility EDF that it has added pound 1.5 bn (Euro 1.7 bn, $ 1.9 bn) to its estimated costs for two new reactors at Hinkley Point C, has led to questions about whether the government should rethink the project, with some politicians calling for it to be abandoned. EDF's announcement came less than two weeks after a report from the Government Accountability Office (GAO) said the government's deal for the two EPR units, now estimated to be costing pound 19.6 bn (Euro 22.3 bn, $ 25.5 bn), has locked consumers into a risky and expensive project with uncertain strategic and economic benefits. The UK's Energy Technologies Institute (ETI), in a report prepared before the EDF announcement, had already said the requirement to improve the predictability and affordability of new nuclear power plants has never been stronger.

GAO report and EDF cost revisions reignite debate on Hinkley Point C

International Nuclear Information System (INIS)

Dalton, David

2017-01-01

The announcement by French state-controlled utility EDF that it has added pound 1.5 bn (Euro 1.7 bn, $ 1.9 bn) to its estimated costs for two new reactors at Hinkley Point C, has led to questions about whether the government should rethink the project, with some politicians calling for it to be abandoned. EDF's announcement came less than two weeks after a report from the Government Accountability Office (GAO) said the government's deal for the two EPR units, now estimated to be costing pound 19.6 bn (Euro 22.3 bn, $ 25.5 bn), has locked consumers into a risky and expensive project with uncertain strategic and economic benefits. The UK's Energy Technologies Institute (ETI), in a report prepared before the EDF announcement, had already said the requirement to improve the predictability and affordability of new nuclear power plants has never been stronger.

Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

International Nuclear Information System (INIS)

Killeen, J.C.

1997-01-01

A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12μm grain size fuel. Both large grain size variants had similar grain sizes around 35μm. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of 85 Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs

Fission gas release during post irradiation annealing of large grain size fuels from Hinkley point B

Energy Technology Data Exchange (ETDEWEB)

Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

1997-08-01

A series of post-irradiation anneals has been carried out on fuel taken from an experimental stringer from Hinkley Point B AGR. The stringer was part of an experimental programme in the reactor to study the effect of large grain size fuel. Three differing fuel types were present in separate pins in the stringer. One variant of large grain size fuel had been prepared by using an MgO dopant during fuel manufactured, a second by high temperature sintering of standard fuel and the third was a reference, 12{mu}m grain size fuel. Both large grain size variants had similar grain sizes around 35{mu}m. The present experiments took fuel samples from highly rated pins from the stringer with local burn-up in excess of 25GWd/tU and annealed these to temperature of up to 1535 deg. C under reducing conditions to allow a comparison of fission gas behaviour at high release levels. The results demonstrate the beneficial effect of large grain size on release rate of {sup 85}Kr following interlinkage. At low temperatures and release rates there was no difference between the fuel types, but at temperatures in excess of 1400 deg. C the release rate was found to be inversely dependent on the fuel grain size. The experiments showed some differences between the doped and undoped large grains size fuel in that the former became interlinked at a lower temperature, releasing fission gas at an increased rate at this temperature. At higher temperatures the grain size effect was dominant. The temperature dependence for fission gas release was determined over a narrow range of temperature and found to be similar for all three types and for both pre-interlinkage and post-interlinkage releases, the difference between the release rates is then seen to be controlled by grain size. (author). 4 refs, 7 figs, 3 tabs.

After Hinkley

International Nuclear Information System (INIS)

Sweet, C.

1990-01-01

An energy consultant to Friends of the Earth and the National Union of Mineworkers (NUM), gave evidence at the Hinkley Inquiry on behalf of the NUM. His evidence compared the economic costs of coal and nuclear generation. Following the revelations by the Central Electricity Generating Board (CEGB) about cost escalation at Sizewell B, he submitted further evidence which showed that the cost of electricity from Sizewell B and the proposed PWR at Hinkley C would be somewhere between 6p/kWh and 10p/kWh. In this article he looks at the role of the public inquiries and summarises the evidence given by the NUM at Hinkley. The importance of the Inquiry is seen to lie in its broader political role. It provided an opportunity for the case against nuclear power to be advanced from a public platform. The establishment have to address that case. Evidence otherwise not available has to be produced, raising the level of debate above the normal level of supply side propaganda. It involved large numbers of people, many in the role of objectors, in active opposition. There is the strategically important time factor - the GEGB's interest in PWRs began in 1972, in 1978 the government agreed to an option for two PWRs. In 1979 this was raised to 10, but only half of one had been built by 1989. (author)

Report of NII investigation into allegations of faulty welding at Hinkley 'B' nuclear power station

International Nuclear Information System (INIS)

1987-01-01

This reports the procedure and findings of the Nuclear Installations Inspectorate's investigation into allegations of welding and radiography malpractice at Hinkley Point-B power station. These concerned welds and their radiographic testing made on pipework carrying water or steam associated with one of the main electricity turbo generators, during construction in 1971. The water or steam is not radioactive and pipe failure would have no nuclear safety significance. Both the Central Electricity Generating Board and the NII investigated the allegations. Both investigations concluded that there was no evidence to support the allegations. (U.K.)

The orificing of once-through boilers for gas-cooled reactors

International Nuclear Information System (INIS)

Collier, J.G.; Whitmarsh-Everiss, M.J.

1986-01-01

The boilers at Hinkley Point B, Hunterston B (which are built to a common design) and Hartlepool and Heysham (which are identical to each other although different from Hinkley B/Hunterston B) are discussed. Heysham II and Torness have boilers developed from the Hinkley B/Hunterston B design. The boiler design is explained and the problems encountered are presented. The study presented in this address highlights the potential sensitivity of the performance of once-through boilers to the following factors - the uniformity of the gas inlet temperature profile, the choice of the operating point on the boiler characteristic (preferably this should be where the performance of the unit is not too sensitive to the primary side temperature and flow distribution); this can be achieved by greater levels of orificing reducing the steady state gains, and the various plant operating states which introduce asymmetries into the boundary conditions for the boiler. A number of computer codes have been used to model boiler performance. Using these, a new design was proposed for boiler feed regulating orifices and these revised orifices were installed during outages on the two Heysham reactors in 1985. (U.K.)

In-core flow measurement and calibration of gags using on-load instrumented stringers in a C.A.G.R. at Hinkley Point 'B'

International Nuclear Information System (INIS)

Harrison, W.E.; Carrick, I.H.

1982-06-01

The initial fuel loading of the first CAGR at Hinkley Point included 5 specially instrumented stringers (OLIS) each containing a flow-measuring venturi and additional thermocouples. Venturi absolute and differential pressures were measured by transducers mounted on the pile-cap. Transducers and thermocouples were routed to a computer/logger and processed into stringer performance data. The venturi was engineered to comply closely with appropriate British Standards but compromises were made to minimise interaction with other functions of the OLIS plug unit, justifying rig calibration of venturis to check for deviation in behaviour. High accuracy and reliability of the flow measuring system were established by thorough commissioning procedures. The transducers were selected for low sensitivity to their operational environment. Nevertheless calibration of all transducers was carried out both in laboratory and in-situ. Errors introduced by signal processing were identified and zero drift monitored. Pipe-runs were scrupulously leak-tested and leak sensitivity was evaluated. After one year re-calibration and recommissioning gave confidence of long term stability. Measurements of stringer behaviour were collected in a series of tests spanning the full range of both the setting of the channel flow control gags and the reactor power. Throughout these tests comprehensive monitoring, with intercalibration between the OLIS and comparison with installed reactor instrumentation has quantified residual error. These measurements were used to check the theoretical model used by the station for channel flow assessment. The excellent agreement obtained justified proceeding to the derivation of a universal gag resistance calibration applying to all power levels. In performance tests aimed at evaluation of overall generating efficiency, the theoretical model was used to make accurate estimates of reactor power and flow which agreed well with estimates based directly on further OLIS

Improved estimates of external gamma dose rates in the environs of Hinkley Point Power Station

International Nuclear Information System (INIS)

Macdonald, H.F.; Thompson, I.M.G.

1988-07-01

The dominant source of external gamma dose rates at centres of population within a few kilometres of Hinkley Point Power Station is the routine discharge of 41-Ar from the 'A' station magnox reactors. Earlier estimates of the 41-Ar radiation dose rates were based upon measured discharge rates, combined with calculations using standard plume dispersion and cloud-gamma integration models. This report presents improved dose estimates derived from environmental gamma dose rate measurements made at distances up to about 1 km from the site, thus minimising the degree of extrapolation introduced in estimating dose rates at locations up to a few kilometres from the site. In addition, results from associated chemical tracer measurements and wind tunnel simulations covering distances up to about 4 km from the station are outlined. These provide information on the spatial distribution of the 41-Ar plume during the initial stages of its dispersion, including effects due to plume buoyancy and momentum and behaviour under light wind conditions. In addition to supporting the methodology used for the 41-Ar dose calculations, this information is also of generic interest in the treatment of a range of operational and accidental releases from nuclear power station sites and will assist in the development and validation of existing environmental models. (author)

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.

Hinkley Point 'C' power station public enquiry: proof of evidence on coal fired power station sites

Energy Technology Data Exchange (ETDEWEB)

Fothergill, S.; Witt, S.

1988-11-01

The Coalfield Communities Campaign (CCC) has argued that if a new base-load power station is required it should be coal-fired rather than nuclear, and that it should use UK coal. Proposals for new power stations at both Hinkley Point and at Fawley have encountered very considerable local and regional opposition, and this is increasingly likely to be the case at many other sites especially in Southern England. In contrast the CCC has sought to demonstrate that its member authorities would generally welcome the development of new coal-fired capacity on appropriate sites within their areas. In particular, this proof establishes that there is a prima facie case for considering three sites - Thorpe Marsh, Hams Hall and Uskmouth - as potential locations for a new large coal-fired power station as an alternative to Hinkley Point C. The relevant local authorities have expressed their willingness to co-operate in more detailed planning or technical investigations to secure a coal-fired power station on these sites. The CCC considers this to be a major and unprecedented offer to the CEGB and its successor bodies, which could greatly speed the development of new power staion capacity and be of considerable economic and social benefit to coalfield communities.

Genetic algorithms and artificial neural networks for loading pattern optimisation of advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Ziver, A.K. E-mail: a.k.ziver@imperial.ac.uk; Pain, C.C; Carter, J.N.; Oliveira, C.R.E. de; Goddard, A.J.H.; Overton, R.S

2004-03-01

A non-generational genetic algorithm (GA) has been developed for fuel management optimisation of Advanced Gas-Cooled Reactors, which are operated by British Energy and produce around 20% of the UK's electricity requirements. An evolutionary search is coded using the genetic operators; namely selection by tournament, two-point crossover, mutation and random assessment of population for multi-cycle loading pattern (LP) optimisation. A detailed description of the chromosomes in the genetic algorithm coded is presented. Artificial Neural Networks (ANNs) have been constructed and trained to accelerate the GA-based search during the optimisation process. The whole package, called GAOPT, is linked to the reactor analysis code PANTHER, which performs fresh fuel loading, burn-up and power shaping calculations for each reactor cycle by imposing station-specific safety and operational constraints. GAOPT has been verified by performing a number of tests, which are applied to the Hinkley Point B and Hartlepool reactors. The test results giving loading pattern (LP) scenarios obtained from single and multi-cycle optimisation calculations applied to realistic reactor states of the Hartlepool and Hinkley Point B reactors are discussed. The results have shown that the GA/ANN algorithms developed can help the fuel engineer to optimise loading patterns in an efficient and more profitable way than currently available for multi-cycle refuelling of AGRs. Research leading to parallel GAs applied to LP optimisation are outlined, which can be adapted to present day LWR fuel management problems.

The Heysham 2 and Torness nuclear power station projects

International Nuclear Information System (INIS)

Cameron, P.J.

1989-01-01

At the beginning of the design of the Heysham 2 and Torness Nuclear Power Stations in the UK, it was agreed that the Hinkley Point B/Hunterston B Advanced Gas Cooled reactor design would be repeated. However, the Hinkley Point B/Hunterston B designs dated from 1964, and inevitably safety requirements had escalated in the intervening years. Furthermore, operating experience gained from Hinkley Point B/Hunterston B showed that it was desirable to make changes to particular features. Design changes were therefore included where it was considered essential to improve performance, to improve safety and to improve engineering. The resulting station designs are described. (author)

Plume tracer experiments at Hinkley Point 'A' [Nuclear Power Station] during 1987

International Nuclear Information System (INIS)

Foster, P.M.

1988-11-01

The results of the first part of a programme of plume dispersion measurements at the Hinkley Point Nuclear Power Station are described. Using SF 6 gas and pyrotechnic smoke tracer techniques developed during an earlier study at Oldbury, measurements of ground level plume behaviour out to about 4 km and elevated plume behaviour out to about 1 km have been made in a series of twelve 1 hour trials and one 15 minute trial. Whereas the Oldbury study considered passive emissions, attention in this study has been focussed on the behaviour of the buoyant shield cooling air emission. Data on plume rise and the degree of plume entrainment by the building wake and on the effects of entrainment and wind meander on plume width and concentration, are presented and discussed in relation to current modelling recommendations. A limited number of 10 minute averaged measurements of plume concentration and 41-Ar decay gamma count were also made at 2 km range and their correlation and variability examined. (author)

The decision on the application to carry out a decommissioning project at Hinkley Point A Power Station under the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999

International Nuclear Information System (INIS)

2003-01-01

European Council Directive 85/337/EEC, as amended by Council Directive 97/1 I/EC, sets out a framework on the assessment of the effects of certain public and private projects on the environment. The Directive is implemented in Great Britain for decommissioning nuclear reactor projects by the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999. The intention of the Directive and Regulations is to involve the public through consultation in considering the potential environmental impacts of a decommissioning project, and to make the decision-making process on granting consent open and transparent. The Regulations require the licensee to undertake an environmental impact assessment, prepare an environmental statement that summarises the environmental effects of the project, and apply to the Health and Safety Executive (HSE) for consent to carry out a decommissioning project. There is an optional stage where the licensee may request from HSE an opinion on what the environmental statement should contain (called a pre-application opinion). The licensee of Hinkley Point A Power Station, Magnox Electric pie, requested a pre-application opinion and provided information in a scoping report in December 2000. HSE undertook a public consultation on the scoping report and provided its pre- application opinion in April 2001. The licensee applied to HSE for consent to carry out a decommissioning project and provided an environmental statement in December 2001. Following a public consultation on the environmental statement, HSE requested further information that was subsequently provided by the licensee. A further public consultation was undertaken on the further information that ended in March 2003. All these public consultations involved around 60 organisations. HSE granted consent to carry out a decommissioning project at Hinkley Point A Power Station under the Regulations in July 2003, and attached conditions to the Consent. HSE took relevant

Equipment and techniques developed for the repair of reactor standpipes

International Nuclear Information System (INIS)

Hartnell, P.; Kirton-Darling, H.F.R.; Walton, D.N.

1993-01-01

In-service cracks had developed in welds in the fabrication at the bottom of the standpipe liner at Hinkley Point-B and Hunterston-B reactors. A joint team investigating these cracks determined that the hot gas flow patterns in the plug unit to standpipe liner annulus were unstable. This had led to thermal stressing and consequent cracking of the standpipe liner welds. In developing a solution consideration was to be given to access down the standpipe for equipment, ambient temperatures, radiation levels and safety considerations. The procedures and equipment developed are described. (author)

Progress with the AGR system

International Nuclear Information System (INIS)

Merrett, D.J.

1984-01-01

The AGR programme was initiated in 1965 with the ordering of the Dungeness 'B' reactor, followed by Hinkley point 'B' (1965), Hunterston 'B' (1968), Hartlepool (1970), Heysham I (1970) and the two latest stations at Heysham II and Torness. The paper reviews the achievements and prospects for the AGR system under 6 topic headings. These include: operational experience at Hinkley Point 'B' and Hunterston'B', commissioning of Dungeness 'B', Hartlepool and Heysham I, Heysham II/Torness design, Heysham II/Torness programme and finally future prospects. (U.K.)

Hinkley Point A gas duct repairs

International Nuclear Information System (INIS)

Curtis, R.F.

1996-01-01

In 1990, routine visual inspection of the Hinckley Point A Reactor 1 pressure vessel gas outlet ducts showed failures in the welded stud bolts retaining the insulation edging strips. Since the ducts are accessible only from within the pressure vessel, a remote repair technique that could be deployed via the vessel stand pipe had to be found. A drawn arc stud welding and work package formerly used at the Oldbury Power Station was modified for the purpose. The only manipulators with sufficient reach and adequate carrying capacity to deploy the package were the Sizewell SNAKES manipulators. One of these was modified to fit the Hinckley reactor and repairs have been successfully carried out. Similar studs on the gas ducts in Reactor 2, are shielded from visual inspection by a Z-clip feature. A technique using pulsed thermography was developed. The studs were heated for a short time at their exposed ends using a prefocused lamp and the heat decay patterns monitored by an infrared camera enabling attached and detached studs to be distinguished. The inspection package was deployed using the SNAKES manipulator again. In both operations, I-Grip computer modelling was used in the design of the package envelope and the deployment routes. (UK)

Power station impacts: socio-economic impact assessment

International Nuclear Information System (INIS)

Glasson, John; Elson, Martin; Barrett, Brendan; Wee, D. Van der

1987-01-01

The aim of this study is to assess the local social and economic impacts of a proposed nuclear power station development at Hinkley Point in Somerset. The proposed development, Hinkley Point C, would be an addition to the existing Hinkley Point A Magnox station, commissioned in 1965, and the Hinkley Point B Advanced Gas Cooled Reactor (AGR) station, commissioned in 1976. It is hoped that the study will be of assistance to the CEGB, the Somerset County and District Councils and other agencies in their studies of the proposed development. In addition, the study seeks to apply and further develop the methodology and results from previous studies by the Power Station Impacts (PSI) team for predicting the social and economic effects of proposed power station developments on their localities. (author)

Construction and commissioning of the Hinkley Point 'B' and Hunterston 'B' boilers

International Nuclear Information System (INIS)

White, D.C.; Holmes, R.L.

1977-01-01

Prior to construction and erection of the reactor boiler units, within the concrete pressure vessel, the units were received from the manufacturing works, and stored in clean and humidity controlled conditions. Because of the loading facilities into the reactor pressure vessel, a very precise erection procedure had to be adhered to. The activities associated with construction of feedwater inlet, superheater steam, and reheated steam inlet and outlet penetrations had to be programmed accordingly. By the very nature of the work load, the time scale involved, and the prime need to maintain the boiler unit materials free of deterioration from atmospheric corrosion after erection, early commissioning of storage systems were implemented; providing wet or dry storage conditions as dictated by the two Generating Boards. Pre-operational commissioning also covered the work of setting up all the steam and water controls, isolating valve systems and the automatic and sequence equipment associated with the feed water controls which are a major design feature of the once through boiler operation. (author)

As the UK searches for post-Brexit investments, why is the new PM stalling Hinkley Point C?

International Nuclear Information System (INIS)

Shepherd, John

2016-01-01

Since voting in a referendum to leave the European Union, the UK has been hard at work on ''shaping the destiny'' of its place in the big wide world outside the family of EU nations - particularly in terms of business and investment. Shortly after this referendum EDF announced the decision to invest in the Hinkley Point C nuclear project. Inexplicably, new UK prime minister Theresa May said there would be no green light or signing of investment guarantees by her new administration, which ''needed more time'' to consider the project. Nuclear power is not the only industry that the UK will need to nurture to make an economic success of the country in the future - but nuclear will be key to the UK's industrial and economic success.

As the UK searches for post-Brexit investments, why is the new PM stalling Hinkley Point C?

Energy Technology Data Exchange (ETDEWEB)

Shepherd, John [nuclear 24, Redditch (United Kingdom)

2016-08-15

Since voting in a referendum to leave the European Union, the UK has been hard at work on ''shaping the destiny'' of its place in the big wide world outside the family of EU nations - particularly in terms of business and investment. Shortly after this referendum EDF announced the decision to invest in the Hinkley Point C nuclear project. Inexplicably, new UK prime minister Theresa May said there would be no green light or signing of investment guarantees by her new administration, which ''needed more time'' to consider the project. Nuclear power is not the only industry that the UK will need to nurture to make an economic success of the country in the future - but nuclear will be key to the UK's industrial and economic success.

Safety design features for current UK advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

1981-01-15

The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

Safety design features for current UK advanced gas-cooled reactors

International Nuclear Information System (INIS)

Yellowlees, J.M.; Cobb, E.C.

1981-01-01

The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

The operator's requirement for in service inspection at Hinkley Point 'B' power station

International Nuclear Information System (INIS)

Outram, J.F.

The subject is discussed under the following heads: the need for reactor internal inspection (safety; commercial interest); the methods available (man access inspection; remote viewing inspection); experience with man access; experience with remote viewing; reactor outage time -comparison of the two methods; contamination control; conclusions. (U.K.)

Middle Pleistocene infill of Hinkley Valley by Mojave River sediment and associated lake sediment: Depositional architecture and deformation by strike-slip faults

Science.gov (United States)

Miller, David; Haddon, Elizabeth; Langenheim, Victoria; Cyr, Andrew J.; Wan, Elmira; Walkup, Laura; Starratt, Scott W.

2018-01-01

Hinkley Valley in the Mojave Desert, near Barstow about 140 km northeast of Los Angeles and midway between Victorville Valley and the Lake Manix basin, contains a thick sedimentary sequence delivered by the Mojave River. Our study of sediment cores drilled in the valley indicates that Hinkley Valley was probably a closed playa basin with stream inflow from four directions prior to Mojave River inflow. The Mojave River deposited thick and laterally extensive clastic wedges originating from the southern valley that rapidly filled much of Hinkley Valley. Sedimentary facies representing braided stream, wetland, delta, and lacustrine depositional environments all are found in the basin fill; in some places, the sequence is greater than 74 m (245 ft) thick. The sediment is dated in part by the presence of the ~631 ka Lava Creek B ash bed low in the section, and thus represents sediment deposition after Victorville basin was overtopped by sediment and before the Manix basin began to be filled. Evidently, upstream Victorville basin filled with sediment by about 650 ka, causing the ancestral Mojave River to spill to the Harper and Hinkley basins, and later to Manix basin.Initial river sediment overran wetland deposits in many places in southern Hinkley Valley, indicating a rapidly encroaching river system. These sediments were succeeded by a widespread lake (“blue” clay) that includes the Lava Creek B ash bed. Above the lake sediment lies a thick section of interlayered stream sediment, delta and nearshore lake sediment, mudflat and/or playa sediment, and minor lake sediment. This stratigraphic architecture is found throughout the valley, and positions of lake sediment layers indicate a successive northward progression in the closed basin. A thin overlapping sequence at the north end of the valley contains evidence for a younger late Pleistocene lake episode. This late lake episode, and bracketing braided stream deposits of the Mojave River, indicate that the river

EDF Energy Nuclear New Build: Lessons Learned in Knowledge Management

International Nuclear Information System (INIS)

Sachar, M.; Borlodan, G.

2016-01-01

Full text: EDF Energy Nuclear New Build (NNB) is building two EPR reactors at Hinkley Point C in Somerset in the United Kingdom that will provide reliable, low carbon electricity to meet approximately 7% of the UK’s electricity needs. The Hinkley Point C project is well advanced. It has achieved planning consent, design approval for the EPR reactor and a nuclear site license. There is a well-developed supply chain with identified preferred bidders who are already heavily involved in construction planning. Training for needed skills is underway and industrial agreements with trade unions are in place. NNB has the unique opportunity to set Knowledge Management behaviours, culture, and standards for the Hinkley Point C project from project inception instead of working to change them, such as on an operational site. (author

Nuclear. Getting out of the EPR deadlock: the EPR's evil star; Under the Flamanville vessel; United-Kingdom: Hinkley Point, stop or more? A nuclear which we will be able to finance

International Nuclear Information System (INIS)

Dupin, Ludovic

2016-01-01

A first article outlines and comments the difficulties faced in the development and construction of the EPR which seems to be born under an evil star, and which resulted in a loss of credibility for the French nuclear industry. The author evokes various steps in the design and test of various components which finally produced important delays and high costs. Moreover, the reactor high power seems not to be adapted to current market status and needs. The reorganisation of the French nuclear sector also produced an unfavourable environment (the dismantling of Areva into two companies is briefly evoked). Now, the objective for EDF is to try to optimize the reactor production and reduce costs by 25 to 30 per cent. A second article addresses the situation of the Flamanville EPR construction which is now six years late. The author proposes an overview of the differences between the various parts and components of the power station: the control room is operational whereas many parts and rooms are still under construction. A new management of construction organisation has been set up. The third article addresses the situation of the British Hinkley Point EPR project. Due to the difficulties met in Finland and in Flamanville, a new financial drift would be a catastrophe, and as many aspects of the project are already well defined, EDF keeps on stating that the decision to build these two EPRs is about to be taken, as it is in fact always delayed. The last article is an interview in which a manager of an important nuclear engineering company comments the role of the French nuclear reactor model in the world, the development of concepts of small modular reactors, and the impact of EPR construction difficulties on the image of the French nuclear industry in the world

Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

International Nuclear Information System (INIS)

Pattinson, A.

2003-01-01

The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

Challenges lie ahead, but Hinkley can be 'on time and within budget'

International Nuclear Information System (INIS)

Kraev, Kamen

2017-01-01

Paris-based engineering services company Assystem is working with EDF to prepare for construction of the UK's first nuclear unit in more than 20 years at Hinkley Point. Christian Jeanneau, the company's senior vice-president, nuclear, spoke to NucNet about what has been done so far and the need to learn lessons from other EPR projects. Assystem has experience with the French PWR fleet, but has also had a presence in the UK for about 40 years with around 1,200 employees in various locations across the country. Other major projects Assystem is involved in are the early development phases of NuGen's AP1000 new-build project for the proposed Moorside project in the UK as well as the United Arab Emirates and the Turkish nuclear new-build programme.

How safe is an AGR

International Nuclear Information System (INIS)

Wilkinson, Max

1987-01-01

The paper concerns the safety of an AGR, in the light of the Chernobyl and Three Mile Island reactor accidents. To assess the safety systems the resources editor of the Financial Times newspaper spent an afternoon trying to do as much damage as possible to one of the Hinkley Point B advanced gas cooled reactors - on the simulator at the Central Electricity Generating Board's training centre at Oldbury. An account of the experience in the nuclear power control room is given. (U.K.)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Advanced gas cooled reactors - Designing for safety

International Nuclear Information System (INIS)

Keen, Barry A.

1990-01-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

Advanced gas cooled reactors - Designing for safety

Energy Technology Data Exchange (ETDEWEB)

Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

1990-07-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

Challenges lie ahead, but Hinkley can be 'on time and within budget'

Energy Technology Data Exchange (ETDEWEB)

Kraev, Kamen [NucNet, Brussels (Belgium)

2017-04-15

Paris-based engineering services company Assystem is working with EDF to prepare for construction of the UK's first nuclear unit in more than 20 years at Hinkley Point. Christian Jeanneau, the company's senior vice-president, nuclear, spoke to NucNet about what has been done so far and the need to learn lessons from other EPR projects. Assystem has experience with the French PWR fleet, but has also had a presence in the UK for about 40 years with around 1,200 employees in various locations across the country. Other major projects Assystem is involved in are the early development phases of NuGen's AP1000 new-build project for the proposed Moorside project in the UK as well as the United Arab Emirates and the Turkish nuclear new-build programme.

Turning points in reactor design

International Nuclear Information System (INIS)

Beckjord, E.S.

1995-01-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

Turning points in reactor design

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-09-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

International Nuclear Information System (INIS)

Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

2003-01-01

A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

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)

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.

Construction of the PCPV's at Hinkley Point 'B' and Hunterston 'B'

International Nuclear Information System (INIS)

Jones, W.C.; Taylor, S.J.

1976-01-01

A general description is given of the PCPVs. An outline of the construction programming is given, highlighting the attention required to the provision of access and the forward planning necessary to reduce interference between the various contractors. The attachment of the bottom liner to the vessel concrete required substantial experimental work and full-scale trials were carried out. In the central zone of the top slab the standpipes are closely pitched with minimum ligament widths between the outside of adjacent standpipes of about 60 mm. Again, full-scale trials were carried out to determine the most satisfactory method of concreting in this area to achieve sound concrete throughout the slab. More detailed information is given on these two items, together with other general construction techniques such as tendon duct fixing, methods of concreting and associated plant, foundation problems encountered and embedment of instrumentation. A brief description is given of the site organisation, temporary works and how some of the day to day problems were met and overcome. (author)

Steam generator materials constraints in UK design gas-cooled reactors

International Nuclear Information System (INIS)

James, D.W.

1988-01-01

A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

She says no!

International Nuclear Information System (INIS)

Carter, M.

1988-01-01

The Public Inquiry into the Central Electricity Generating Board's proposal to build a pressurized water nuclear power plant at Hinkley in Somerset started in October 1988. The opposition to the proposals comes from two main organisations COLA and SHE. COLA (The Consortium of Opposing Local Authorities) has a Pound 1 million budget and is opposed to a PWR in Somerset. SHE (the Stop Hinkley Expansion Campaign) objects not only to the PWR proposal but to the whole idea of nuclear power. The Town and Country Planning Association also opposes Hinkley Point C, especially with plans to privatise the electricity supply industry now under discussion. COLA has 30 fundamental objections to Hinkley Point C - economic, social, environmental and aesthetic. But both COLA and SHE have an extra argument not used at the Sizewell-B Inquiry. This is the ''Chernobyl Factor''. All the opposition groupings claim that public opinion is also against the proposal. In spite of the opposition, the decision about whether or not to go ahead with Hinkley Point C may well be made by the Minister involved, so that it would be a political decision. (U.K.)

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

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

High-Temperature Gas-Cooled Test Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Reactor coolant flow measurements at Point Lepreau

International Nuclear Information System (INIS)

Brenciaglia, G.; Gurevich, Y.; Liu, G.

1996-01-01

The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

Solution of the reactor point kinetics equations by MATLAB computing

Directory of Open Access Journals (Sweden)

Singh Sudhansu S.

2015-01-01

Full Text Available The numerical solution of the point kinetics equations in the presence of Newtonian temperature feedback has been a challenging issue for analyzing the reactor transients. Reactor point kinetics equations are a system of stiff ordinary differential equations which need special numerical treatments. Although a plethora of numerical intricacies have been introduced to solve the point kinetics equations over the years, some of the simple and straightforward methods still work very efficiently with extraordinary accuracy. As an example, it has been shown recently that the fundamental backward Euler finite difference algorithm with its simplicity has proven to be one of the most effective legacy methods. Complementing the back-ward Euler finite difference scheme, the present work demonstrates the application of ordinary differential equation suite available in the MATLAB software package to solve the stiff reactor point kinetics equations with Newtonian temperature feedback effects very effectively by analyzing various classic benchmark cases. Fair accuracy of the results implies the efficient application of MATLAB ordinary differential equation suite for solving the reactor point kinetics equations as an alternate method for future applications.

A plan for study of hexavalent chromium, CR(VI) in groundwater near a mapped plume, Hinkley, California, 2016

Science.gov (United States)

Izbicki, John A.; Groover, Krishangi D.

2016-01-22

The Pacific Gas and Electric Company (PG&E) Hinkley compressor station, in the Mojave Desert 80 miles northeast of Los Angeles, is used to compress natural gas as it is transported through a pipeline from Texas to California. Between 1952 and 1964, cooling water used at the compressor station was treated with a compound containing chromium to prevent corrosion. After cooling, the wastewater was discharged to unlined ponds, resulting in contamination of soil and groundwater in the underlying alluvial aquifer (Lahontan Regional Water Quality Control Board, 2013). Since 1964, cooling-water management practices have been used that do not contribute chromium to groundwater.In 2007, a PG&E study of the natural background concentrations of hexavalent chromium, Cr(VI), in groundwater estimated average concentrations in the Hinkley area to be 1.2 micrograms per liter (μg/L), with a 95-percent upper-confidence limit of 3.1 μg/L (CH2M-Hill, 2007). The 3.1 μg/L upper-confidence limit was adopted by the Lahontan Regional Water Quality Control Board (RWQCB) as the maximum background concentration used to map the plume extent. In response to criticism of the study’s methodology, and an increase in the mapped extent of the plume between 2008 and 2011, the Lahontan RWQCB (Lahontan Regional Water Quality Control Board, 2012) agreed that the 2007 PG&E background-concentration study be updated.The purpose of the updated background study is to evaluate the presence of natural and man-made Cr(VI) near Hinkley, Calif. The study also is to estimate natural background Cr(VI) concentrations in the aquifer upgradient and downgradient from the mapped Cr(VI) contamination plume, as well as in the plume and near its margins. The study was developed by the U.S. Geological Survey (USGS) in collaboration with a technical working group (TWG) composed of community members, the Independent Review Panel (IRP) Manager (Project Navigator, Ltd.), the Lahontan RWQCB, PG&E, and consultants for PG&E.&E.

What will we do with the low level waste from reactor decommissioning?

International Nuclear Information System (INIS)

Meehan, A. R.; Wilmott, S.; Crockett, G.; Watt, N. R.

2008-01-01

The decommissioning of the UK's Magnox reactor sites will produce large volumes of low level waste (LLW) arisings. The vast majority of this waste takes the form of concrete, building rubble and redundant plant containing relatively low levels of radioactivity. Magnox Electric Ltd (Magnox) is leading a strategic initiative funded by the Nuclear Decommissioning Authority (NDA) to explore opportunities for the disposal of such waste to suitably engineered facilities that might be located on or adjacent to the site of waste arising, if appropriate and subject to regulatory acceptance and stakeholder views. The strategic issues surrounding this initiative are described along with an update of progress with stakeholder consultations in relation to the proposed licensing of the first such facility at Hinkley Point A, which could be viewed as a test case for the development of similar disposal facilities at other nuclear sites in England and Wales. (authors)

A two-point kinetic model for the PROTEUS reactor

International Nuclear Information System (INIS)

Dam, H. van.

1995-03-01

A two-point reactor kinetic model for the PROTEUS-reactor is developed and the results are described in terms of frequency dependent reactivity transfer functions for the core and the reflector. It is shown that at higher frequencies space-dependent effects occur which imply failure of the one-point kinetic model. In the modulus of the transfer functions these effects become apparent above a radian frequency of about 100 s -1 , whereas for the phase behaviour the deviation from a point model already starts at a radian frequency of 10 s -1 . (orig.)

Optimal configuration of spatial points in the reactor cell

International Nuclear Information System (INIS)

Bosevski, T.

1968-01-01

Optimal configuration of spatial points was chosen in respect to the total number needed for integration of reactions in the reactor cell. Previously developed code VESTERN was used for numerical verification of the method on a standard reactor cell. The code applies the collision probability method for calculating the neutron flux distribution. It is shown that the total number of spatial points is twice smaller than the respective number of spatial zones needed for determination of number of reactions in the cell, with the preset precision. This result shows the direction for further condensing of the procedure for calculating the space-energy distribution of the neutron flux in a reactors cell [sr

Method of nuclear reactor control using a variable temperature load dependent set point

International Nuclear Information System (INIS)

Kelly, J.J.; Rambo, G.E.

1982-01-01

A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow

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)

INDIAN POINT REACTOR STARTUP AND PERFORMANCE

Energy Technology Data Exchange (ETDEWEB)

Deddens, J. C.; Batch, M. L.

1963-09-15

The testing program for the Indian Point Reactor is discussed. The thermal and hydraulic evaluation of the primary coolant system is discussed. Analyses of fuel loading and initial criticality, measurement of operating coefficients of reactivity, control rod group reactivity worths, and xenon evaluation are presented. (R.E.U.)

Study of the stochastic point reactor kinetic equation

International Nuclear Information System (INIS)

Gotoh, Yorio

1980-01-01

Diagrammatic technique is used to solve the stochastic point reactor kinetic equation. The method gives exact results which are derived from Fokker-Plank theory. A Green's function dressed with the clouds of noise is defined, which is a transfer function of point reactor with fluctuating reactivity. An integral equation for the correlation function of neutron power is derived using the following assumptions: 1) Green's funntion should be dressed with noise, 2) The ladder type diagrams only contributes to the correlation function. For a white noise and the one delayed neutron group approximation, the norm of the integral equation and the variance to mean-squared ratio are analytically obtained. (author)

Fast neutron reactors: the safety point of view

International Nuclear Information System (INIS)

Laverie, M.; Avenas, M.

1984-01-01

All versions of nuclear reactors present favourable and unfavourable characteristics from the point of view of safety. The safety of the installations is obtained by making efforts to utilize in the best possible way those which are favourable and by taking proper steps in the face of those which are unfavourable. The present article shows how this general principle has been applied as regards the fast neutron reactors of integrated design which have been developped in France, taking into account the specific features of this version. A qualitative method to compare the safety of this version with that of pressurized water reactors which has been widely put to the test commercially all over the world is presented. These analyses make, generally speaking, several positive characteristics stand out for these fast neutron reactors from the safety aspects [fr

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.)

An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

International Nuclear Information System (INIS)

Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

2013-01-01

Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

From the nuclear world, no.5

International Nuclear Information System (INIS)

Anon.

2016-01-01

This document gathers information from around the world and concerning nuclear industry. The most relevant is the following. The British government has given its agreement for the construction of 2 EPR by EDF Energy at Hinkley Point. The Hinkley Point project will generate more than 1700 jobs in France. The EPR being built in Finland will operate in 2018. The electrical car is sustainable only in the countries where low-carbon electricity production is important which is the case of France thanks to nuclear energy. In the framework of the ICERR cooperation, 2 research reactors of CEA: Isis (Saclay) and RJH (being built at Cadarache) and their experimental facilities will be used by Slovenia, Tunisia and Morocco through joint research projects. In China 6 AP1000 units will be built on 3 sites in the Yangtze region. China and Turkey have signed an agreement concerning the organisation of nuclear safety authority. In Turkey a site for a 3. nuclear power plant is being selected. (A.C.)

Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

International Nuclear Information System (INIS)

Mitari, O.; Hirose, A.; Skarsgard, H.M.

1989-01-01

In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

Operating point considerations for the Reference Theta-Pinch Reactor (RTPR)

International Nuclear Information System (INIS)

Krakowski, R.A.; Miller, R.L.; Hagenson, R.L.

1976-01-01

Aspects of the continuing engineering design-point reassessment and optimization of the Reference Theta-Pinch Reactor (RTPR) are discussed. An updated interim design point which achieves a favorable energy balance and involves relaxed technological requirements, which nonetheless satisfy more rigorous physics and engineering constraints, is presented

Fractional neutron point kinetics equations for nuclear reactor dynamics

International Nuclear Information System (INIS)

Espinosa-Paredes, Gilberto; Polo-Labarrios, Marco-A.; Espinosa-Martinez, Erick-G.; Valle-Gallegos, Edmundo del

2011-01-01

The fractional point-neutron kinetics model for the dynamic behavior in a nuclear reactor is derived and analyzed in this paper. The fractional model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional order, acting as exponent of the relaxation time, to obtain the best representation of a nuclear reactor dynamics. The physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. The numerical approximation to the solution of the fractional neutron point kinetics model, which can be represented as a multi-term high-order linear fractional differential equation, is calculated by reducing the problem to a system of ordinary and fractional differential equations. The numerical stability of the fractional scheme is investigated in this work. Results for neutron dynamic behavior for both positive and negative reactivity and for different values of fractional order are shown and compared with the classic neutron point kinetic equations. Additionally, a related review with the neutron point kinetics equations is presented, which encompasses papers written in English about this research topic (as well as some books and technical reports) published since 1940 up to 2010.

Compact reversed-field pinch reactors (CRFPR): sensitivity study and design-point determination

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1982-07-01

If the costing assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves overly optimistic, approaches that promise considerably increased system power density and reduced mass utilization will be required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique, magnetic topology associated with the Reversed-Field Pinch (RFP), the compact reactor embodiment for this approach is particularly attractive from the viewpoint of low-field resistive coils operating with Ohmic losses that can be made small relative to the fusion power. A comprehensive system model is developed and described for a steady-state, compact RFP reactor (CRFPR). This model is used to select a unique cost-optimized design point that will be used for a conceptual engineering design. The cost-optimized CRFPR design presented herein would operate with system power densities and mass utilizations that are comparable to fission power plants and are an order of magnitude more favorable than the conventional approaches to magnetic fusion power. The sensitivity of the base-case design point to changes in plasma transport, profiles, beta, blanket thickness, normal vs superconducting coils, and fuel cycle (DT vs DD) is examined. The RFP approach is found to yield a point design for a high-power-density reactor that is surprisingly resilient to changes in key, but relatively unknown, physics and systems parameters

The Hinkley Point decision: An analysis of the policy process

International Nuclear Information System (INIS)

Thomas, Stephen

2016-01-01

In 2006, the British government launched a policy to build nuclear power reactors based on a claim that the power produced would be competitive with fossil fuel and would require no public subsidy. A decade later, it is not clear how many, if any, orders will be placed and the claims on costs and subsidies have proved false. Despite this failure to deliver, the policy is still being pursued with undiminished determination. The finance model that is now proposed is seen as a model other European countries can follow so the success or otherwise of the British nuclear programme will have implications outside the UK. This paper contends that the checks and balances that should weed out misguided policies, have failed. It argues that the most serious failure is with the civil service and its inability to provide politicians with high quality advice – truth to power. It concludes that the failure is likely to be due to the unwillingness of politicians to listen to opinions that conflict with their beliefs. Other weaknesses include the lack of energy expertise in the media, the unwillingness of the public to engage in the policy process and the impotence of Parliamentary Committees. - Highlights: •Britain's nuclear power policy is failing due to high costs and problems of finance. •This has implications for European countries who want to use the same financing model. •The continued pursuit of a failing policy is due to poor advice from civil servants. •Lack of expertise in the media and lack of public engagement have contributed. •Parliamentary processes have not provided proper critical scrutiny.

End point control of an actinide precipitation reactor

International Nuclear Information System (INIS)

Muske, K.R.

1997-01-01

The actinide precipitation reactors in the nuclear materials processing facility at Los Alamos National Laboratory are used to remove actinides and other heavy metals from the effluent streams generated during the purification of plutonium. These effluent streams consist of hydrochloric acid solutions, ranging from one to five molar in concentration, in which actinides and other metals are dissolved. The actinides present are plutonium and americium. Typical actinide loadings range from one to five grams per liter. The most prevalent heavy metals are iron, chromium, and nickel that are due to stainless steel. Removal of these metals from solution is accomplished by hydroxide precipitation during the neutralization of the effluent. An end point control algorithm for the semi-batch actinide precipitation reactors at Los Alamos National Laboratory is described. The algorithm is based on an equilibrium solubility model of the chemical species in solution. This model is used to predict the amount of base hydroxide necessary to reach the end point of the actinide precipitation reaction. The model parameters are updated by on-line pH measurements

Decontamination of the Douglas Point reactor, May 1983

International Nuclear Information System (INIS)

Lesurf, J.E.; Stepaniak, R.; Broad, L.G.; Barber, W.G.

1983-01-01

The Douglas Point reactor primary heat transport system including the fuel, was successfully decontaminated by the CAN-DECON process in 1975. A second decontamination, also using the CAN-DECON process, was successfully performed in May 1983. This paper outlines the need for the decontamination, the process used, the results obtained, and the benefits to the station maintenance and operation

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

Point design for deuterium-deuterium compact reversed-field pinch reactors

International Nuclear Information System (INIS)

Dabiri, A.E.; Dobrott, D.R.; Gurol, H.; Schnack, D.D.

1984-01-01

A deuterium-deuterium (D-D) reversed-field pinch (RFP) reactor may be made comparable in size and cost to a deuterium-tritium (D-T) reactor at the expense of high-thermal heat load to the first wall. This heat load is the result of the larger percentage of fusion power in charged particles in the D-D reaction as compared to the D-T reaction. The heat load may be reduced by increasing the reactor size and hence the cost. In addition to this ''degraded'' design, the size may be kept small by means of a higher heat load wall, or by means of a toroidal divertor, in which case most of the heat load seen by the wall is in the form of radiation. Point designs are developed for these approaches and cost studies are performed and compared with a D-T reactor. The results indicate that the cost of electricity of a D-D RFP reactor is about20% higher than a D-T RFP reactor. This increased cost could be offset by the inherent safety features of the D-D fuel cycle

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

Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

International Nuclear Information System (INIS)

Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

2010-01-01

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

Torness. Keep it green

International Nuclear Information System (INIS)

Flood, M.

1979-01-01

The subject is discussed under the following headings: Torness foolishness (concerning the steps being taken towards constructing a nuclear power station at Torness Point in East Lothian, Scotland); the case against Torness (a summary); is Torness necessary (discusses the present, planned and forecast positions of electricity supply and demand in Scotland); how much will Torness cost, and who will pay; the local impact of Torness (on employment, infrastructure, amenity and future prospects); how reliable is the Advanced Gas Cooled Reactor (the choice of the AGR type reactor and discussions on the 'Hinkley Point incident' and the 'Hunterston incident'); how safe is nuclear power (discussion on AGR programme, nuclear accidents, sabotage, war, nuclear waste, plutonium, secrecy and security); public opposition to Torness; what can be done. (U.K.)

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Development and operation of the Above Dome Inspection Rig (ADIR)

International Nuclear Information System (INIS)

Dickson, R.P.; Moorby, J.

1984-01-01

Hinkley Point 'B' is developing its remote inspection equipment in order to be able to inspect reactor internals adequately without manned vessel entry. The Above Dome Inspection Rig has been built to allow a number of inspection systems to be introduced and operated within the reactor. The ability to introduce, use and remove inspection equipment without the necessity to lift the rig from the reactor is a vital feature in the speed achieved in completing inspections quickly. Television was selected for the A.D.M. because it has significant advantages in terms of operational convenience. However the quality of image obtained in terms of information available compares unfavourably with photography. The sharpness of a photographic image is largely dictated by the chemical structure of the emulsion, whereas video is limited by the picture line structure and bandwidth. The need for a photographic system for in reactor use is therefore essential for high definition inspection requirements. The first inspection system that has been developed for the ADIR is the Telefilm camera. It consists of a Hasselblad photographic camera using an Insight television camera looking through its viewfinder. The characteristics of television and photography have been combined. (author)

Real-time simulation of response to load variation for a ship reactor based on point-reactor double regions and lumped parameter model

Energy Technology Data Exchange (ETDEWEB)

Wang Qiao; Zhang De [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China); Chen Wenzhen, E-mail: Cwz2@21cn.com [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China); Chen Zhiyun [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China)

2011-05-15

Research highlights: > We calculate the variation of main parameters of the reactor core by the Simulink. > The Simulink calculation software (SCS) can deal well with the stiff problem. > The high calculation precision is reached with less time, and the results can be easily displayed. > The quick calculation of ship reactor transient can be achieved by this method. - Abstract: Based on the point-reactor double regions and lumped parameter model, while the nuclear power plant second loop load is increased or decreased quickly, the Simulink calculation software (SCS) is adopted to calculate the variation of main physical and thermal-hydraulic parameters of the reactor core. The calculation results are compared with those of three-dimensional simulation program. It is indicated that the SCS can deal well with the stiff problem of the point-reactor kinetics equations and the coupled problem of neutronics and thermal-hydraulics. The high calculation precision can be reached with less time, and the quick calculation of parameters of response to load disturbance for the ship reactor can be achieved. The clear image of the calculation results can also be displayed quickly by the SCS, which is very significant and important to guarantee the reactor safety operation.

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.)

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.)

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)

Numerical simulation of stochastic point kinetic equation in the dynamical system of nuclear reactor

International Nuclear Information System (INIS)

Saha Ray, S.

2012-01-01

Highlights: ► In this paper stochastic neutron point kinetic equations have been analyzed. ► Euler–Maruyama method and Strong Taylor 1.5 order method have been discussed. ► These methods are applied for the solution of stochastic point kinetic equations. ► Comparison between the results of these methods and others are presented in tables. ► Graphs for neutron and precursor sample paths are also presented. -- Abstract: In the present paper, the numerical approximation methods, applied to efficiently calculate the solution for stochastic point kinetic equations () in nuclear reactor dynamics, are investigated. A system of Itô stochastic differential equations has been analyzed to model the neutron density and the delayed neutron precursors in a point nuclear reactor. The resulting system of Itô stochastic differential equations are solved over each time-step size. The methods are verified by considering different initial conditions, experimental data and over constant reactivities. The computational results indicate that the methods are simple and suitable for solving stochastic point kinetic equations. In this article, a numerical investigation is made in order to observe the random oscillations in neutron and precursor population dynamics in subcritical and critical reactors.

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

Space nuclear reactor concepts for avoidance of a single point failure

International Nuclear Information System (INIS)

El-Genk, M. S.

2007-01-01

This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

Review of Kaganove's solution for the reactor point kinetics equations

International Nuclear Information System (INIS)

Couto, R.T.; Santo, A.C.F. de.

1993-09-01

A review of Kaganove's method for the reactor point kinetics equations solution is performed. This was method chosen to calculate the power in ATR, a computer program for the analysis of reactivity transients. The reasons for this choice and the adaptation of the method to the purposes of ATR are presented. (author)

Application of the fractional neutron point kinetic equation: Start-up of a nuclear reactor

International Nuclear Information System (INIS)

Polo-Labarrios, M.-A.; Espinosa-Paredes, G.

2012-01-01

Highlights: ► Neutron density behavior at reactor start up with fractional neutron point kinetics. ► There is a relaxation time associated with a rapid variation in the neutron flux. ► Physical interpretation of the fractional order is related with non-Fickian effects. ► Effect of the anomalous diffusion coefficient and the relaxation time is analyzed. ► Neutron density is related with speed and duration of the control rods lifting. - Abstract: In this paper we present the behavior of the variation of neutron density when the nuclear reactor power is increased using the fractional neutron point kinetic (FNPK) equation with a single-group of delayed neutron precursor. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and its physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. We analyzed the case of increase the nuclear reactor power when reactor is cold start-up which is a process of inserting reactivity by lifting control rods discontinuously. The results show that for short time scales of the start-up the neutronic density behavior with FNPK shows sub-diffusive effects whose absorption are government by control rods velocity. For large times scale, the results shows that the classical equation of the neutron point kinetics over predicted the neutron density regarding to FNPK.

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

Comparative study on nutrient removal of agricultural non-point source pollution for three filter media filling schemes in eco-soil reactors.

Science.gov (United States)

Du, Fuyi; Xie, Qingjie; Fang, Longxiang; Su, Hang

2016-08-01

Nutrients (nitrogen and phosphorus) from agricultural non-point source (NPS) pollution have been increasingly recognized as a major contributor to the deterioration of water quality in recent years. The purpose of this article is to investigate the discrepancies in interception of nutrients in agricultural NPS pollution for eco-soil reactors using different filling schemes. Parallel eco-soil reactors of laboratory scale were created and filled with filter media, such as grit, zeolite, limestone, and gravel. Three filling schemes were adopted: increasing-sized filling (I-filling), decreasing-sized filling (D-filling), and blend-sized filling (B-filling). The systems were intermittent operations via simulated rainstorm runoff. The nutrient removal efficiency, biomass accumulation and vertical dissolved oxygen (DO) distribution were defined to assess the performance of eco-soil. The results showed that B-filling reactor presented an ideal DO for partial nitrification-denitrification across the eco-soil, and B-filling was the most stable in the change of bio-film accumulation trends with depth in the three fillings. Simultaneous and highest removals of NH4(+)-N (57.74-70.52%), total nitrogen (43.69-54.50%), and total phosphorus (42.50-55.00%) were obtained in the B-filling, demonstrating the efficiency of the blend filling schemes of eco-soil for oxygen transfer and biomass accumulation to cope with agricultural NPS pollution.

Cost-constrained design point for the Reversed-Field Pinch Reactor (RFPR)

International Nuclear Information System (INIS)

Hagenson, R.L.; Krakowski, R.A.

1978-01-01

A broad spectrum of Reversed-Field Pinch Reactor (RFPR) operating modes are compared on an economics basis. An RFPR with superconducting coils and an air-core poloidal field transformer optimizes to give a minimum cost system when compared to normal-conducting coils and the iron-core transformer used in earlier designs. An interim design is described that exhibits a thermally stable, unrefueled, 21 s burn (burnup 50 percent) with an energy containment time equal to 200 times the Bohm time, which is consistent with present-day tokamak experiments. This design operates near the minimum energy state (THETA = B/sub THETA/(r/sub w/)/[B/sub z/] = 2.0 and F = B/sub z/(r/sub w/)/[B/sub z/] = 1.0 from the High Beta Model) of the RFP configuration. This cost-optimized design produces a reactor of 1.5-m minor radius and 12.8-m major radius, that generates 1000 MWe (net) with a recirculating power fraction of 0.15 at a direct capital cost of 970 $/kWe

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

The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

International Nuclear Information System (INIS)

Miller, R.L.

1989-01-01

The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

Iron-free moving coil high temperature displacement transducer

Energy Technology Data Exchange (ETDEWEB)

Grindrod, A

1976-07-01

A unique, iron free, moving coil linear displacement transducer system is described, which is suitable for continuously monitoring linear movements, at varying temperatures up to 750/sup 0/C, in operational nuclear reactors. Although this device has been primarily developed for Advanced Gas Cooled Reactor Systems, it also has uses where long term measurements on conventional high temperature plant are required. Furthermore it could be particularly useful in material creep laboratories where precise linear changes in specimen length need to be monitored at elevated temperatures, over several years. Since individual transducer installations demand specific mounting arrangements to suit particular component geometries, evaluations have been made only on standard operational modules or capsules which are designed for containment in a range of housing or fixtures to suit particular applications. The behaviour of these devices has been studied at temperatures up to 750/sup 0/C for periods of over 10,000 h. An evaluation is also included of a commercially designed sensor assembly employing the same principle, for monitoring the boiler-shield wall movement at Hinkley Point 'B' AGR Station.

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.)

PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

International Nuclear Information System (INIS)

Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

1990-01-01

1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

Manufacture of steam generator units and components for the AGR power stations at Heysham II and Torness

International Nuclear Information System (INIS)

Glasgow, J.R.; Parkin, K.

1984-01-01

The current AGR Steam Generator is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley B/Hunterston B AGR power stations. In this paper a brief outline of the evolution of the steam generator design from the earlier gas cooled reactor stations is presented. A description of the main items of fabrication development is given. The production facilities for the manufacture of the units are described. Reference is also made to some of the work on associated components. The early experience on the construction site of installation of the steam generators is briefly outlined. (author)

Manufacture of steam generator units and components for the AGR power stations at Heysham II and Torness

Energy Technology Data Exchange (ETDEWEB)

Glasgow, J R; Parkin, K [N.E.I. Nuclear Systems Ltd., Gateshead, Tyne and Wear (United Kingdom)

1984-07-01

The current AGR Steam Generator is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley B/Hunterston B AGR power stations. In this paper a brief outline of the evolution of the steam generator design from the earlier gas cooled reactor stations is presented. A description of the main items of fabrication development is given. The production facilities for the manufacture of the units are described. Reference is also made to some of the work on associated components. The early experience on the construction site of installation of the steam generators is briefly outlined. (author)

Aging management program of the reactor building concrete at Point Lepreau Generating Station

Science.gov (United States)

Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

2011-04-01

In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

A new integral method for solving the point reactor neutron kinetics equations

International Nuclear Information System (INIS)

Li Haofeng; Chen Wenzhen; Luo Lei; Zhu Qian

2009-01-01

A numerical integral method that efficiently provides the solution of the point kinetics equations by using the better basis function (BBF) for the approximation of the neutron density in one time step integrations is described and investigated. The approach is based on an exact analytic integration of the neutron density equation, where the stiffness of the equations is overcome by the fully implicit formulation. The procedure is tested by using a variety of reactivity functions, including step reactivity insertion, ramp input and oscillatory reactivity changes. The solution of the better basis function method is compared to other analytical and numerical solutions of the point reactor kinetics equations. The results show that selecting a better basis function can improve the efficiency and accuracy of this integral method. The better basis function method can be used in real time forecasting for power reactors in order to prevent reactivity accidents.

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

Fully 3D printed integrated reactor array for point-of-care molecular diagnostics.

Science.gov (United States)

Kadimisetty, Karteek; Song, Jinzhao; Doto, Aoife M; Hwang, Young; Peng, Jing; Mauk, Michael G; Bushman, Frederic D; Gross, Robert; Jarvis, Joseph N; Liu, Changchun

2018-06-30

Molecular diagnostics that involve nucleic acid amplification tests (NAATs) are crucial for prevention and treatment of infectious diseases. In this study, we developed a simple, inexpensive, disposable, fully 3D printed microfluidic reactor array that is capable of carrying out extraction, concentration and isothermal amplification of nucleic acids in variety of body fluids. The method allows rapid molecular diagnostic tests for infectious diseases at point of care. A simple leak-proof polymerization strategy was developed to integrate flow-through nucleic acid isolation membranes into microfluidic devices, yielding a multifunctional diagnostic platform. Static coating technology was adopted to improve the biocompatibility of our 3D printed device. We demonstrated the suitability of our device for both end-point colorimetric qualitative detection and real-time fluorescence quantitative detection. We applied our diagnostic device to detection of Plasmodium falciparum in plasma samples and Neisseria meningitides in cerebrospinal fluid (CSF) samples by loop-mediated, isothermal amplification (LAMP) within 50 min. The detection limits were 100 fg for P. falciparum and 50 colony-forming unit (CFU) for N. meningitidis per reaction, which are comparable to that of benchtop instruments. This rapid and inexpensive 3D printed device has great potential for point-of-care molecular diagnosis of infectious disease in resource-limited settings. Copyright © 2018 Elsevier B.V. All rights reserved.

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

Science.gov (United States)

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Aging management program of the reactor building concrete at Point Lepreau Generating Station

Directory of Open Access Journals (Sweden)

Gendron T.

2011-04-01

Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

Nucleaire et Energies Nr 68 - June 2016

International Nuclear Information System (INIS)

Lenail, Bernard; Ducroux, Guy; Seyve, Claude; Simonnet, Jacques; Dupont, Jean-Francois; Salanave, Jean-Luc; Justin, Francois; Greneche, Dominique; Lepine, Gerard; Raisonnier, Daniele

2016-06-01

After a first article which comments the situation of EDF which is a serious matter of concern, notably due to the current context for the energy sector and market, and which notably comments the issue of the Hinkley Point project, an article proposes an overview of recent evolutions (oil price recent evolutions, perspectives for gas and coal, the abundance and availability of hydroelectric energy). Three articles concern nuclear activities: recent events about reactors (EU policy, decisions, projects and events in Switzerland, Sweden, UK about Hinkley Point, Fessenheim, Chernobyl, Poland, Europe about reactor lifetime extension, Russia, in Saudi Arabia and South Korea about the Korean APR-1400 technology, in Bangladesh, in India, Japan and USA), about the back-end of the fuel cycle and decommissioning activities (brief focus on the Cigeo project, foreign storages, discussion of the warehousing-transport issue, dismantling activities in France, recycling in France, Russia, China, Taiwan and the UK, evolution of the MFFF plant in Savannah River), and a comment of the main qualities of nuclear energy. Articles comment issues concerning the relationship between nuclear and society with a comment on the French approach (nuclear and precaution principle, anti-nuclear position of French media, crucial role of operators for nuclear safety), and a comment on the evolution of human lifetime expectation statistics in France and its possible relationship with technology when taking accidents related to energy sources into account. An article analyses the German situation with respect to energy. A last article briefly discusses some terminological confusion between energy and electricity, and between produced energy and available power

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)

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.

The application of polynomial chaos methods to a point kinetics model of MIPR: An Aqueous Homogeneous Reactor

International Nuclear Information System (INIS)

Cooling, C.M.; Williams, M.M.R.; Nygaard, E.T.; Eaton, M.D.

2013-01-01

Highlights: • A point kinetics model for the Medical Isotope Production Reactor is formulated. • Reactivity insertions are simulated using this model. • Polynomial chaos is used to simulate uncertainty in reactor parameters. • The computational efficiency of polynomial chaos is compared to that of Monte Carlo. -- Abstract: This paper models a conceptual Medical Isotope Production Reactor (MIPR) using a point kinetics model which is used to explore power excursions in the event of a reactivity insertion. The effect of uncertainty of key parameters is modelled using intrusive polynomial chaos. It is found that the system is stable against reactivity insertions and power excursions are all bounded and tend towards a new equilibrium state due to the negative feedbacks inherent in Aqueous Homogeneous Reactors (AHRs). The Polynomial Chaos Expansion (PCE) method is found to be much more computationally efficient than that of Monte Carlo simulation in this application

Arnott untangles Webb

International Nuclear Information System (INIS)

Arnott, Don.

1989-01-01

Dr Webb, a former nuclear scientist, reactor physicist and reactor operator now concentrates his efforts on studying the capacity for catastrophic explosions (fires and steam explosions) in PWR type reactors. He gave evidence to the Hinkley Public Inquiry, presenting evidence which analysed the accident hazards of the PWR. This profile shows Dr Webb to be pro-safety rather than anti-reactor. He would support a reactor design which fails safe whatever happens. His talents as a writer and academic are renowned. (U.K.)

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

Application of point kinetic model in the study of fluidized bed reactor dynamic

International Nuclear Information System (INIS)

Borges, Volnei; Vilhena, Marco Tullio de; Streck, Elaine E.

1995-01-01

In this work the dynamical behavior of the fluidized bed nuclear reactor is analysed. The main goal consist to study the effect of the acceleration term in the point kinetic equations. Numerical simulations are reported considering constant acceleration. (author). 7 refs, 4 figs

POINT 2011: ENDF/B-VII.1 Beta2 Temperature Dependent Cross Section Library

Energy Technology Data Exchange (ETDEWEB)

Cullen, D E

2011-04-07

This report is one in the series of 'POINT' reports that over the years have presented temperature dependent cross sections for the then current version of ENDF/B. In each case I have used my personal computer at home and publicly available data and codes. I have used these in combination to produce the temperature dependent cross sections used in applications and presented in this report. I should mention that today anyone with a personal computer can produce these results. The latest ENDF/B-VII.1 beta2 data library was recently and is now freely available through the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. This release completely supersedes all preceding releases of ENDF/B. As distributed the ENDF/B-VII.1 data includes cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in our applications the ENDF/B-VII.1 library has been processed into cross sections at eight neutron reactor like temperatures, between 0 and 2100 Kelvin, in steps of 300 Kelvin (the exception being 293.6 Kelvin, for exact room temperature at 20 Celsius). It has also been processed to five astrophysics like temperatures, 1, 10, 100 eV, 1 and 10 keV. For reference purposes, 300 Kelvin is approximately 1/40 eV, so that 1 eV is approximately 12,000 Kelvin. At each temperature the cross sections are tabulated and linearly interpolable in energy. All results are in the computer independent ENDF-6 character format [R2], which allows the data to be easily transported between computers. In its processed form the POINT 2011 library is approximately 16 gigabyte in size and is distributed on one compressed DVDs (see, below for the details of the contents of each DVD).

An efficient technique for the point reactor kinetics equations with Newtonian temperature feedback effects

International Nuclear Information System (INIS)

Nahla, Abdallah A.

2011-01-01

Highlights: → An efficient technique for the nonlinear reactor kinetics equations is presented. → This method is based on Backward Euler or Crank Nicholson and fundamental matrix. → Stability of efficient technique is defined and discussed. → This method is applied to point kinetics equations of six-groups of delayed neutrons. → Step, ramp, sinusoidal and temperature feedback reactivities are discussed. - Abstract: The point reactor kinetics equations of multi-group of delayed neutrons in the presence Newtonian temperature feedback effects are a system of stiff nonlinear ordinary differential equations which have not any exact analytical solution. The efficient technique for this nonlinear system is based on changing this nonlinear system to a linear system by the predicted value of reactivity and solving this linear system using the fundamental matrix of the homogenous linear differential equations. The nonlinear point reactor kinetics equations are rewritten in the matrix form. The solution of this matrix form is introduced. This solution contains the exponential function of a variable coefficient matrix. This coefficient matrix contains the unknown variable, reactivity. The predicted values of reactivity in the explicit form are determined replacing the exponential function of the coefficient matrix by two kinds, Backward Euler and Crank Nicholson, of the rational approximations. The nonlinear point kinetics equations changed to a linear system of the homogenous differential equations. The fundamental matrix of this linear system is calculated using the eigenvalues and the corresponding eigenvectors of the coefficient matrix. Stability of the efficient technique is defined and discussed. The efficient technique is applied to the point kinetics equations of six-groups of delayed neutrons with step, ramp, sinusoidal and the temperature feedback reactivities. The results of these efficient techniques are compared with the traditional methods.

The achievement of on-load refuelling at Heysham 2 and Torness AGRs

International Nuclear Information System (INIS)

Sterland, P.R.; MacPherson, D.

1995-01-01

Heysham 2 and Torness are the last of the advanced gas cooled reactors (AGR) constructed for the Central Electricity Generating Board and the South of Scotland Electricity Board. They are now operated by Nuclear Electric (NE) and Scottish Nuclear Limited (SNL) respectively. They were designed for on load refuelling at high power and being based on the design used for Hinkley Point B and Hunterston B, have benefited from experience at these stations. This paper examines the analysis work which was carried out in order to provide a sound, long term, mainly probabilistic safety case which supports on load refuelling. A significant problem has been that the design reliability of the microprocessor based control and protection systems, used for the fuel route, could not be justified and a number of changes to the overall protection system have had to be introduced to compensate for this. The resultant safety case complies with the stringent safety standards adopted by NE and SNL and accepted by the UK safety authority, the Nuclear Installations Inspectorate (NII). (author)

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

International Nuclear Information System (INIS)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

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

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

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

Methods for solving the stochastic point reactor kinetic equations

International Nuclear Information System (INIS)

Quabili, E.R.; Karasulu, M.

1979-01-01

Two new methods are presented for analysis of the statistical properties of nonlinear outputs of a point reactor to stochastic non-white reactivity inputs. They are Bourret's approximation and logarithmic linearization. The results have been compared with the exact results, previously obtained in the case of Gaussian white reactivity input. It was found that when the reactivity noise has short correlation time, Bourret's approximation should be recommended because it yields results superior to those yielded by logarithmic linearization. When the correlation time is long, Bourret's approximation is not valid, but in that case, if one can assume the reactivity noise to be Gaussian, one may use the logarithmic linearization. (author)

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.)

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)

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

Protection set-points lines for the reactor core and considerations about power distribution and peak factors

International Nuclear Information System (INIS)

Furieri, E.B.

1981-01-01

In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

Point Lepreau generating station

International Nuclear Information System (INIS)

Ganong, G.H.D.; Strang, A.E.; Gunter, G.E.; Thompson, T.S.

Point Lepreau-1 reactor is a 600 MWe generating station expected to be in service by October 1979. New Brunswick is suffering a 'catch up' phenomenon in load growth and needs to decrease dependence on foreign oil. The site is on salt water and extensive study has gone into corrosion control. Project management, financing and scheduling have unique aspects. (E.C.B.)

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)

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

International Nuclear Information System (INIS)

Palma, Daniel A.P.; Martinez, Aquilino S.; Goncalves, Alessandro C.

2009-01-01

The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

Analytical solution of point kinetics equations for linear reactivity variation during the start-up of a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Palma, Daniel A.P. [CEFET QUIMICA de Nilopolis/RJ, 21941-914 Rio de Janeiro (Brazil)], E-mail: agoncalves@con.ufrj.br; Martinez, Aquilino S.; Goncalves, Alessandro C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)

2009-09-15

The analytical solution of point kinetics equations with a group of delayed neutrons is useful in predicting the variation of neutron density during the start-up of a nuclear reactor. In the practical case of an increase of nuclear reactor power resulting from the linear insertion of reactivity, the exact analytical solution cannot be obtained. Approximate solutions have been obtained in previous articles, based on considerations that need to be verifiable in practice. In the present article, an alternative analytic solution is presented for point kinetics equations in which the only approximation consists of disregarding the term of the second derivative for neutron density in relation to time. The results proved satisfactory when applied to practical situations in the start-up of a nuclear reactor through the control rods withdraw.

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

The achievement of on-load refuelling at Heysham 2 and Torness AGRs

International Nuclear Information System (INIS)

Sterland, P.R.; MacPherson, D.

1995-01-01

Heysham 2 and Torness are the last of the advanced gas cooled reactors constructed for the Central Electricity Generating Board and the South of Scotland Electricity Board. They are now operated by Nuclear Electric (NE) and Scottish Nuclear Limited (SNL), respectively. They were designed for on-load refuelling at high power and, being based on the design used for Hinkley Point B and Hunterston B, have benefited from experience at these stations. This paper examines the analysis work which was carried out in order to provide a sound, long-term, mainly probabilistic safety case which supports on-load refuelling. A significant problem has been that the design reliability of the microprocessor-based control and protection systems, used for the fuel route, could not be justified, and a number of changes to the overall protection system have had to be introduced to compensate for this. The safety assessment has covered the fuelling machine pressure boundary, the hoist and fuel assembly components, possible faults and the protection which is necessary to prevent them, the fuel and fuelling machine cooling requirements, the mechanical and radiological consequences of dropping fuel, and the effect of refuelling transients on the reactor and conventional plant. The resultant safety case complies with the stringent safety standards adopted by NE and SNL and accepted by the UK safety authority, the Nuclear Installations Inspectorate (NII). (author)

The extension of the SWS period or CANDU reactors with particular reference to Douglas Point

International Nuclear Information System (INIS)

Bennett, C.R.

1985-01-01

The foregoing approach to the determination of the fate of a concrete containment building is worth much consideration. The expenditure of $10 8 or its escalated equivalent is too much to pay for the probable saving of fraction of a statistical life. The unquestioning adoption of the dogma of reactor dismantlement displays a complete misunderstanding of the numerics of ''risk'', even the place of reactor dismantling in the spectrum of nuclear risk. The position of the risk of reactor dismantling is more than an order of magnitude lower than the former of these. The most altruistic criterion for any engineering activity is the achievement of the greatest expected net benefit (or the least expected net detriment) when all the consequences of the activity are taken into account. As has been shown this criterion leads to the conclusion that, at least in CANDU reactors and particularly Douglas Point, there is apparently no reason why the S.W.S. period should not be extended indefinitely

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

Contour analysis of steady state tokamak reactor performance

International Nuclear Information System (INIS)

Devoto, R.S.; Fenstermacher, M.E.

1990-01-01

A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

A core-monitoring based methodology for predictions of graphite weight loss in AGR moderator bricks

Energy Technology Data Exchange (ETDEWEB)

McNally, K., E-mail: kevin.mcnally@hsl.gsi.gov.uk [Health and Safety Laboratory, Harpur Hill, Buxton, Derbyshire SK17 9JN (United Kingdom); Warren, N. [Health and Safety Laboratory, Harpur Hill, Buxton, Derbyshire SK17 9JN (United Kingdom); Fahad, M.; Hall, G.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, University of Manchester, Manchester M13 9PL (United Kingdom)

2017-04-01

Highlights: • A statistically-based methodology for estimating graphite density is presented. • Graphite shrinkage is accounted for using a finite element model. • Differences in weight loss forecasts were found when compared to the existing model. - Abstract: Physically based models, resolved using the finite element (FE) method are often used to model changes in dimensions and the associated stress fields of graphite moderator bricks within a reactor. These models require inputs that describe the loading conditions (temperature, fluence and weight loss ‘field variables’), and coded relationships describing the behaviour of graphite under these conditions. The weight loss field variables are calculated using a reactor chemistry/physics code FEAT DIFFUSE. In this work the authors consider an alternative data source of weight loss: that from a longitudinal dataset of density measurements made on small samples trepanned from operating reactors during statutory outages. A nonlinear mixed-effect model is presented for modelling the age and depth-related trends in density. A correction that accounts for irradiation-induced dimensional changes (axial and radial shrinkage) is subsequently applied. The authors compare weight loss forecasts made using FEAT DIFFUSE with those based on an alternative statistical model for a layer four moderator brick for the Hinkley Point B, Reactor 3. The authors compare the two approaches for the weight loss distribution through the brick with a particular focus on the interstitial keyway, and for the average (over the volume of the brick) weight loss.

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.

Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor

Directory of Open Access Journals (Sweden)

Nowak Tomasz Karol

2014-06-01

Full Text Available This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.

Bubble-point and dew-point equation for binary refrigerant mixture R22-R142b

Energy Technology Data Exchange (ETDEWEB)

Liancheng Tan; Zhongyou Zhao; Yonghong Duan (Xi' an Jiaotong Univ., Xi' an (China). Dept. of Power Machinery Engineering)

1992-01-01

A bubble-point and dew-point equation (in terms either of temperature or of pressure is suggested for the refrigerant mixture R22-R142b), which is regarded as one of the alternatives to R12. This equation has been examined with experimental data. A modified Rackett equation for the calculation of the bubble-point volume is also proposed. Compared with the experimental data, the rms errors in the calculated values of the bubble-point temperature, the dew-point temperature, and the bubble-point volume are 1.093%, 0.947%, and 1.120%, respectively. The calculation covers a wide range of temperatures and pressures, even near the critical point. It is shown how the equations are extrapolated to calculate other binary refrigerant mixtures. (author)

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)

UABUC - Single energy point model burnup computer code for water reactors

International Nuclear Information System (INIS)

El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

1981-01-01

UABUC is a single energy point reactor burnup computer program in FORTRAN language. The program calculates the change in the isotopic composition of the uranium fuel as a function of irradiation time with all its associated quantities such as the average point flux, the conversion ratio, macroscopic fuel cross sections, and the point reactivity profile. A step-wise time analytical solution was developed for the nonlinear first order burnup differential equations. The ''Westcott'' convention of the effective cross sections was used except for plutonium-240 and uranium-238. For plutonium-240, an effective microscopic cross section was derived from the direct physical arguments taking into account the selfshielding effect of plutonium-240 as well as the 1 ev. resonance absorption. For uranium-238, an effective cross section, reflecting the effect of fast fission and resonance absorption was used. The fission products were treated in the three groups with 50, 300, and 800 barns. The yields in the groups were treated as functions of the type of fissionable nuclides, the effective neutron temperature, and the epithermal index. Xenon-135 and Samarium-149 were treated separately as functions of irradiation time. (author)

Theory of fluctuations and parametric noise in a point nuclear reactor model

International Nuclear Information System (INIS)

Rodriguez, M.A.; San Miguel, M.; Sancho, J.M.

1984-01-01

We present a joint description of internal fluctuations and parametric noise in a point nuclear reactor model in which delayed neutrons and a detector are considered. We obtain kinetic equations for the first moments and define effective kinetic parameters which take into account the effect of parametric Gaussian white noise. We comment on the validity of Langevin approximations for this problem. We propose a general method to deal with weak but otherwise arbitrary non-white parametric noise. Exact kinetic equations are derived for Gaussian non-white noise. (author)

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-02-01

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF₂) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

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

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

International Nuclear Information System (INIS)

Park, T-K.; Jung, S-H.

1996-01-01

During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

A small modular fast reactor as starting point for industrial deployment of fast reactors

International Nuclear Information System (INIS)

Chang, Yoon I.; Lo Pinto, Pierre; Konomura, Mamoru

2006-01-01

The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO 2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

Determination of the protection set-points lines for the Angra-1 reactor core

International Nuclear Information System (INIS)

Furieri, E.B.

1980-03-01

In this work several thermo-hidraulic calculation were performed to obtain Protection set-points lines for the Angra-1 reactor core in order to compare with the values presented by the vendor in the FSAR. These lines are the locus of points where DNBR min = 1,3 and power = 1,18 x P nominal as a function of ΔT m and T m , the temperature difference and the average coolant temperature between hot and cold legs. A computation scheme was developed using COBRA-IIIF as a subroutine of a new main program and adding new subroutines in order to obtain the desired DNBR. The solution is obtained through a convergentce procedure using parameters estimated in a sensivity study. (author) [pt

Analytic method study of point-reactor kinetic equation when cold start-up

International Nuclear Information System (INIS)

Zhang Fan; Chen Wenzhen; Gui Xuewen

2008-01-01

The reactor cold start-up is a process of inserting reactivity by lifting control rod discontinuously. Inserting too much reactivity will cause short-period and may cause an overpressure accident in the primary loop. It is therefore very important to understand the rule of neutron density variation and to find out the relationships among the speed of lifting control rod, and the duration and speed of neutron density response. It is also helpful for the operators to grasp the rule in order to avoid a start-up accident. This paper starts with one-group delayed neutron point-reactor kinetics equations and provides their analytic solution when reactivity is introduced by lifting control rods discontinuously. The analytic expression is validated by comparison with practical data. It is shown that the analytic solution agrees well with numerical solution. Using this analytical solution, the relationships among neutron density response with the speed of lifting control rod and its duration are also studied. By comparing the results with those under the condition of step inserted reactivity, useful conclusions are drawn

Development of a point-kinetic verification scheme for nuclear reactor applications

Energy Technology Data Exchange (ETDEWEB)

Demazière, C., E-mail: demaz@chalmers.se; Dykin, V.; Jareteg, K.

2017-06-15

In this paper, a new method that can be used for checking the proper implementation of time- or frequency-dependent neutron transport models and for verifying their ability to recover some basic reactor physics properties is proposed. This method makes use of the application of a stationary perturbation to the system at a given frequency and extraction of the point-kinetic component of the system response. Even for strongly heterogeneous systems for which an analytical solution does not exist, the point-kinetic component follows, as a function of frequency, a simple analytical form. The comparison between the extracted point-kinetic component and its expected analytical form provides an opportunity to verify and validate neutron transport solvers. The proposed method is tested on two diffusion-based codes, one working in the time domain and the other working in the frequency domain. As long as the applied perturbation has a non-zero reactivity effect, it is demonstrated that the method can be successfully applied to verify and validate time- or frequency-dependent neutron transport solvers. Although the method is demonstrated in the present paper in a diffusion theory framework, higher order neutron transport methods could be verified based on the same principles.

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J.

1982-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

Comparison of apparent diffusion coefficients (ADCs) between two-point and multi-point analyses using high-B-value diffusion MR imaging

International Nuclear Information System (INIS)

Kubo, Hitoshi; Maeda, Masayuki; Araki, Akinobu

2001-01-01

We evaluated the accuracy of calculating apparent diffusion coefficients (ADCs) using high-B-value diffusion images. Echo planar diffusion-weighted MR images were obtained at 1.5 tesla in five standard locations in six subjects using gradient strengths corresponding to B values from 0 to 3000 s/mm 2 . Estimation of ADCs was made using two methods: a nonlinear regression model using measurements from a full set of B values (multi-point method) and linear estimation using B values of 0 and max only (two-point method). A high correlation between the two methods was noted (r=0.99), and the mean percentage differences were -0.53% and 0.53% in phantom and human brain, respectively. These results suggest there is little error in estimating ADCs calculated by the two-point technique using high-B-value diffusion MR images. (author)

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

World must build two atomic reactors each day the next hundred years. [Summary of and commentary on book, 'Mankind at the Turning Point'

Energy Technology Data Exchange (ETDEWEB)

1974-07-24

In summarizing and commenting on the ideas presented in Mesarovic and Pestel's book ''Mankind at the Turning Point'' it is pointed out that the global energy crisis makes comprehensive long-term planning a necessity. Assuming, optimistically, that nuclear power alone is able to supply the total projected energy demand in 100 years, it is stated that this will require 3000 nuclear power stations, each with 8 fast breeder reactors, totally 100 GW(t). This means a net rate of construction of four reactors per week, which again means allowing for a 30-year life, two reactors per day, every day, for the next hundred years. Fueling of these reactors will require the production and transport of 15 x 10/sup 6/ kg of /sup 239/Pu per year. It is therefore obvious that the energy crisis is not only a technological, but also a political, social, and even psychological problem.

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

The asymptotic behaviour of a critical point reactor in the absence of a controller

International Nuclear Information System (INIS)

Bansal, N.K.; Borgwaldt, H.

1976-11-01

A method is presented to calculate the first and second moments of neutron and precursor populations for a critical reactor system described by point kinetic equations and possessing inherent reactivity fluctuations. The equations have been linearised on the assumption that the system has a large average neutron population and that the amplitude of reactivity fluctuations is sufficiently small. The reactivity noise is assumed to be band limited white with a corner frequency higher than all the time constants of the system. Explicit expressions for the exact time development of the moments have been obtained for the case of a reactor without reactivity feedback and with one group of delayed neutrons. It is found that the expected values of the neutron and delayed neutron precursor numbers tend asymptotically to stationary values, whereas the mean square deviations increase linearly with time at an extremely low rate. (orig.) [de

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

Station blackout and public confidence: a cautionary tale

International Nuclear Information System (INIS)

Cave, L.

1990-01-01

The recent ''station blackout'' (ie loss of on-site and off-site AC power) incidents at the Vogtle PWR in the US and Hinkley Point B AGR in the Uk have led to further public concern about the safety of nuclear power, even though in each case the actual increase in the chance of an accident leading to a release of radioactivity to the environment was negligible. The industry may be wise to invest precautionary measures to reduce the frequency of such incidents and to increase public confidence. (author)

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

Measurement techniques for AGR circulators in a full-density rig

International Nuclear Information System (INIS)

Watson, I.; Wilson, R.R.

1977-01-01

Safety and reliability are the most important factors of a nuclear power plant. This applies in particular to the circulators used to drive the high-density CO 2 around the reactor core and boiler circuits. Under operating conditions, very high sound-pressure levels are generated which could excite components and cause possible fatigue failures. Failures of this type were experienced on the original axial blowers for the Hinkley 'A' Magnox reactor and, following this, a stringent test plan was specified for the AGR circulators. The present paper describes some of the techniques used to measure strain, sound and vibration on circulators in a full-density rig. This rig reproduces the actual reactor working conditions of 300 0 C and 4.1 MN m -2 with gas velocities up to 120 m s -1 . Under these conditions sound-pressure levels of up to 172 dB are generated. This programme of circulator testing has continued for the past 10 years. During this period many obstacles and difficulties were encountered. Some of these problems, together with the solutions found, are discussed. (author)

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.)

Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-12-01

Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

International Nuclear Information System (INIS)

Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

2015-01-01

Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

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

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.)

The cost of SMRs. How Rolls Royce aims to compete with wind and solar

Energy Technology Data Exchange (ETDEWEB)

Dalton, David [NucNet The Independent Global Nuclear News Agency, Brussels (Belgium)

2017-12-15

The UK nuclear industry is hoping that claims by Rolls-Royce that small modular reactor (SMR) projects could deliver electricity for a similar cost to offshore wind will provide much-needed impetus to government plans for the country to develop a ''best value'' SMR and put it into commercial operation by the end of the next decade. Rolls-Royce and its consortium partners, including Amec Foster Wheeler, Arup, Laing O'Rourke and Nuvia, say the UK SMR they are developing could produce energy for as low as pound 60 (Euro 66, $ 79) per MWh, which is competitive against wind and solar and significantly lower than the pound 92.50 per MWh agreed by the government and project developer EDF for the new Hinkley Point C nuclear station.

S.I. 1987 No. 2182, The Electricity Generating Stations and Overhead Lines (Inquiries Procedure) Rules 1987

International Nuclear Information System (INIS)

1987-01-01

These Rules, which came into force on 14 January 1988, make new provision for the procedure for any public inquiry held pursuant to Section 34 of the Electricity Act 1957 in relation to applications for consent to construct or extend a generating station (including nuclear stations). The Rules were made pursuant to Section 11 of the Tribunals and Inquiries Act 1971. They revoke the previous Electricity Generating Stations and Overhead Line (Inquiries Procedures) Rules 1981. These new Rules cover the same topics as the previous Rules but aim to shorten the potential length and thus cost of inquiries. They will apply to the Inquiry to be held into the application by the Central Electricity Generating Board to build a pressurised water reactor at Hinkley Point in Somerset. (NEA) [fr

The cost of SMRs. How Rolls Royce aims to compete with wind and solar

International Nuclear Information System (INIS)

Dalton, David

2017-01-01

The UK nuclear industry is hoping that claims by Rolls-Royce that small modular reactor (SMR) projects could deliver electricity for a similar cost to offshore wind will provide much-needed impetus to government plans for the country to develop a ''best value'' SMR and put it into commercial operation by the end of the next decade. Rolls-Royce and its consortium partners, including Amec Foster Wheeler, Arup, Laing O'Rourke and Nuvia, say the UK SMR they are developing could produce energy for as low as pound 60 (Euro 66, $ 79) per MWh, which is competitive against wind and solar and significantly lower than the pound 92.50 per MWh agreed by the government and project developer EDF for the new Hinkley Point C nuclear station.

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

NJOY processed multigroup library for fast reactor applications and point data library for MCNP - Experience and validation

International Nuclear Information System (INIS)

Kim Jung-Do; Gil Choong-Sup

1996-01-01

JEF-1-based 50-group cross section library for fast reactor applications and point data library for continuous-energy Monte Carlo code MCNP have been generated using NJOY91.38 system. They have been examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 8 small fast critical assemblies. (author). 9 refs, 2 figs, 10 tabs

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.)

World must build two atomic reactors each day the next hundred years. [Summary of and commentary on book, 'Mankind at the Turning Point'

Energy Technology Data Exchange (ETDEWEB)

1974-07-24

In summarizing and commenting on the ideas presented in Mesarovic and Pestel's book ''Mankind at the Turning Point'' it is pointed out that the global energy crisis makes comprehensive long-term planning a necessity. Assuming, optimistically, that nuclear power alone is able to supply the total projected energy demand in 100 years, it is stated that this will require 3000 nuclear power stations, each with 8 fast breeder reactors, totally 100 GW(t). This means a net rate of construction of four reactors per week, which again means allowing for a 30-year life, two reactors per day, every day, for the next hundred years. Fueling of these reactors will require the production and transport of 15 x 10/sup 6/ kg of /sup 239/Pu per year. It is therefore obvious that the energy crisis is not only a technological, but also a political, social, and even psychological problem.

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

International Nuclear Information System (INIS)

Boustani, Ehsan; Amirkabir University of Technology, Tehran; Khakshournia, Samad

2016-01-01

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

Energy Technology Data Exchange (ETDEWEB)

Boustani, Ehsan [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.; Khakshournia, Samad [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.

2016-12-15

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Effect of incisor inclination changes on cephalometric points a and b

International Nuclear Information System (INIS)

Hassan, S.; Shaikh, A.; Fida, M.

2015-01-01

The position of cephalometric points A and B are liable to be affected by alveolar remodelling caused by orthodontic tooth movement during incisor retraction. This study was conducted to evaluate the change in positions of cephalometric points A and B in sagittal and vertical dimensions due to change in incisor inclinations. Methods: Total sample of 31 subjects were recruited into the study. The inclusion criteria were extraction of premolars in upper and lower arches, completion of growth and orthodontic treatment. The exclusion criteria were patients with craniofacial anomalies and history of orthodontic treatment. By superimposition of pre and post treatment tracings, various linear and angular parameters were measured. Various tests and multiple linear regression analysis were performed to determine changes in outcome variables. Statistically significant p-value was <0.05. Results:One-sample t-test showed that change in position of only point A was statistically significant which was 1.61mm (p<0.01) in sagittal direction and 1.49mm (p<0.01) in vertical direction. Multiple linear regression analysis showed that if we retrocline upper incisor by 100, the point A will move superiorly by 0.6mm. Conclusions: Total change in the position of point A is in a downward and forward direction. Total Change in upper incisors inclinations causes change in position of point A only in vertical direction. (author)

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)

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.)

A comparison of mghr prescription to doses at points A and B in intracavitary radiotherapy of cervix cancer

International Nuclear Information System (INIS)

Park, C.I.; Ha, S.W.; Kang, W.S.

1981-01-01

The 42 patients with carcinoma of the cervix, performed intracavitary radiotherapy, were analysed the doses at points A and B comparing to the mghr prescription. The doses at points A and B were calculated by PC-12 computer planning system. Correlation coefficiency between doses at points A and B and the mghr prescription are 0.82 (p<0.001) and 0.90 (p<0.001) respectively. The slope of the point A line is 0.70 and the slope of the point B is 0.21. Therefore, the dose at point A is approximately 3/4 the mghr prescription and the dose at point B is approximately 1/4 the mghr prescription. (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.)

J-R Fracture Resistance of SA533 Gr.B-Cl.1 Steel for Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

A rolled plate might show different mechanical behaviors from a forging, even though they contain same chemical compositions. Furthermore, it is known that the fracture behavior of a rolled plate is very sensitive to material orientation comparing to a forging. In this study, the J-R fracture resistances of SA533 Gr.B-Cl.1 plate were measured at reactor operating temperature and the material orientation sensitivity was discussed. The decrease of fracture resistance of this kind of low alloy steel at an elevated temperature is known as the effect of dynamic strain aging (DSA). It was attributed to that the carbides and grains elongated to primary rolling direction, so that the aspect ratio of carbides and grains in the specimen with T-L orientation is larger. Generally, the hard second phase could take a roll of trigger point of unstable fracture. It is needed that the fracture surfaces of the tested specimens to be examined profoundly.

Dynamic analysis of multiple nuclear-coupled boiling channels based on a multi-point reactor model

International Nuclear Information System (INIS)

Lee, J.D.; Pan Chin

2005-01-01

This work investigates the non-linear dynamics and stabilities of a multiple nuclear-coupled boiling channel system based on a multi-point reactor model using the Galerkin nodal approximation method. The nodal approximation method for the multiple boiling channels developed by Lee and Pan [Lee, J.D., Pan, C., 1999. Dynamics of multiple parallel boiling channel systems with forced flows. Nucl. Eng. Des. 192, 31-44] is extended to address the two-phase flow dynamics in the present study. The multi-point reactor model, modified from Uehiro et al. [Uehiro, M., Rao, Y.F., Fukuda, K., 1996. Linear stability analysis on instabilities of in-phase and out-of-phase modes in boiling water reactors. J. Nucl. Sci. Technol. 33, 628-635], is employed to study a multiple-channel system with unequal steady-state neutron density distribution. Stability maps, non-linear dynamics and effects of major parameters on the multiple nuclear-coupled boiling channel system subject to a constant total flow rate are examined. This study finds that the void-reactivity feedback and neutron interactions among subcores are coupled and their competing effects may influence the system stability under different operating conditions. For those cases with strong neutron interaction conditions, by strengthening the void-reactivity feedback, the nuclear-coupled effect on the non-linear dynamics may induce two unstable oscillation modes, the supercritical Hopf bifurcation and the subcritical Hopf bifurcation. Moreover, for those cases with weak neutron interactions, by quadrupling the void-reactivity feedback coefficient, period-doubling and complex chaotic oscillations may appear in a three-channel system under some specific operating conditions. A unique type of complex chaotic attractor may evolve from the Rossler attractor because of the coupled channel-to-channel thermal-hydraulic and subcore-to-subcore neutron interactions. Such a complex chaotic attractor has the imbedding dimension of 5 and the

Intracavitary dosimetry: a comparison of MGHR prescription to doses at points A and B in cervical cancer

International Nuclear Information System (INIS)

Cunningham, D.E.; Stryker, J.A.; Velkley, D.E.; Chung, C.K.

1981-01-01

This study, involving 77 patients with carcinoma of the cervix, compares the doses at points A and B with the milligram-hour (mg-h) prescription for the intracavitary use of the Fletcher-Suit after loading applicators. The doses at points A and B were computer calculated. A linear least-square regression analysis was used to compare the two sets of data. Correlation coefficients between doses at points A and B and the mg-h prescription are 0.84 (p < 0.001) and 0.88 (p < 0.001) respectively. The slope of the point A line is 0.78 and the slope of the point B line is 0.24. Therefore, for purposes of a nominal comparison, the dose at point A is approximately 3/4 the mg-h prescription; the dose at point B is approximately 1/4 the mg-h prescription. The limitations and significance of the comparison of the two approaches to intracavitary dosimetry is discussed

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

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)

To benefit from the market opportunities, the French nuclear sector must accelerate its change

International Nuclear Information System (INIS)

Therond-Koos, C.

2016-01-01

An independent think-tank gives its meaning on the future of nuclear energy and on the assets of the French nuclear industry to obtain contracts and tenders in international markets. Nuclear power appears to be a necessary passage for a quick transition towards a no-carbon emitting power production. Between 2006 and 2015 the construction of about 80 reactors was launched and IAEA expects a growth rate of nuclear power production somewhere between +50 and +100% by 2030. 6 countries: United States, Russia, France, Japan, China and South-Korea have developed a nuclear industry and are able to export reactor technology. Concerning reactor construction, France's share of the world market is about 10% and France ranks third behind Russia (33%) and American-Japanese companies (33%). China will be soon a reactor supplier. On the nuclear fuel business, AREVA is also a major player. At the world scale only a few companies operate in the reactor construction field because of technological, industrial but also regulation challenges. The Hinkley Point C project is crucial for the French nuclear industry because of the attractiveness of the contract and mainly because it is an opportunity to show the ability of the French nuclear enterprises to end projects on time and it may be considered a reference to win other export contracts. (A.C.)

Innovative reactor core: potentialities and design

International Nuclear Information System (INIS)

Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

2010-01-01

Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

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

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

32 CFR Appendix B to Part 290 - DCAA's FOIA Points of Contact

Science.gov (United States)

2010-07-01

... (CONTINUED) FREEDOM OF INFORMATION ACT PROGRAM DEFENSE CONTRACT AUDIT AGENCY (DCAA) FREEDOM OF INFORMATION ACT PROGRAM Pt. 290, App. B Appendix B to Part 290—DCAA's FOIA Points of Contact (Regional Offices.... Pacific Ocean and Asian Islands. Asia except the Middle East. Australia. Georgia DCAA Eastern Regional...

Pilot program: NRC severe reactor accident incident response training manual. Overview and summary of major points

International Nuclear Information System (INIS)

McKenna, T.J.; Martin, J.A. Jr.; Giitter, J.G.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Watkins

1987-02-01

Overview and Summary of Major Points is the first in a series of volumes that collectively summarize the U.S. Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assessment. Other volumes in the series are: Volume 2-Severe Reactor Accident Overview; Volume 3- Response of Licensee and State and Local Officials; Volume 4-Public Protective Actions-Predetermined Criteria and Initial Actions; Volume 5 - U.S. Nuclear Regulatory Commission. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. The volumes have been organized into these training modules to accommodate the scheduling and duty needs of participating NRC staff. Each volume is accompanied by an appendix of slides that can be used to present this material

The prototype fast reactor

International Nuclear Information System (INIS)

Broomfield, A.M.

1985-01-01

The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

Big rock point restoration project BWR major component removal, packaging and shipping - planning and experience

International Nuclear Information System (INIS)

Milner, T.; Dam, S.; Papp, M.; Slade, J.; Slimp, B.; Nurden, P.

2001-01-01

The Big Rock Point boiling water reactor (BWR) at Charlevoix, MI was permanently shut down on August 29th 1997. In 1999 BNFL Inc.'s Reactor Decommissioning Group (RDG) was awarded a contract by Consumers Energy (CECo) for the Big Rock Point (BRP) Major Component Removal (MCR) project. BNFL Inc. RDG has teamed with MOTA, Sargent and Lundy and MDM Services to plan and execute MCR in support of the facility restoration project. The facility restoration project will be completed by 2005. Key to the success of the project has been the integration of best available demonstrated technology into a robust and responsive project management approach, which places emphasis on safety and quality assurance in achieving project milestones linked to time and cost. To support decommissioning of the BRP MCR activities, a reactor vessel (RV) shipping container is required. Discussed in this paper is the design and fabrication of a 10 CFR Part 71 Type B container necessary to ship the BRP RV. The container to be used for transportation of the RV to the burial site was designed as an Exclusive Use Type B package for shipment and burial at the Barnwell, South Carolina (SC) disposal facility. (author)

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)

Subcutaneous injection of 99mTc pertechnetate at acupuncture points K-3 and B-60

International Nuclear Information System (INIS)

Wu Chung-Chieng; Jong Shiang-Bin; Lin Chun-Ching; Chen Min-Fen; Chen Jong-Rern; Chung Chieng.

1990-01-01

The acupuncture points are known to be morphologically related to the nerves and vessels. Yet the physiological role of blood vessels in the formation of acupuncture points remains unknown. With subcutaneous injection of 99m Tc pertechnetate at the acupuncture points K-3 and B-60 and with intra-acupuncture point injection of 99m Tc pertechnetate at K-3 and B-60, a lower-limb venography like what was obtained by intravenous injection of 99m Tc macroaggregated albumin was demonstrated in the present study. It seems that some acupuncture points do play a role in drainage of tissue fluid from soft tissue into the veins. (author)

The EPR reactor

International Nuclear Information System (INIS)

Lacoste, A.C.; Dupuy, Ph.; Gupta, O.; Perez, J.R.; Emond, D.; Cererino, G.; Rousseau, J.M.; Jeffroy, F.; Evrard, J.M.; Seiler, J.M.; Azarian, G.; Chaumont, B.; Dubail, A.; Fischer, M.; Tiippana, P.; Hyvarinen, J.; Zaleski, C.P.; Meritet, S.; Iglesias, F.; Vincent, C.; Massart, S.; Graillat, G.; Esteve, B.; Mansillon, Y.; Gatinol, C.; Carre, F.

2005-01-01

This document reviews economical and environmental aspects of the EPR project. The following topics are discussed: role and point of view of the French Nuclear Safety Authority on EPR, control of design and manufacturing of EPR by the French Nuclear Safety Authority, assessment by IRSN of EPR safety, research and development in support of EPR, STUK safety review of EPR design, standpoint on EPR, the place of EPR in the French energy policy, the place of EPR in EDF strategy, EPR spearhead of nuclear rebirth, the public debate, the local stakes concerning the building of EPR in France at Flamanville (Manche) and the research on fourth generation reactors. (A.L.B.)

41 CFR Appendix A to Subpart B of... - 3-Key Points and Principles

Science.gov (United States)

2010-07-01

... Principles A Appendix A to Subpart B of Part 102 Public Contracts and Property Management Federal Property.... B, App. A Appendix A to Subpart B of Part 102-3—Key Points and Principles This appendix provides... principles that may be applied to situations not covered elsewhere in this subpart. The guidance follows: Key...

Critical point predication device

International Nuclear Information System (INIS)

Matsumura, Kazuhiko; Kariyama, Koji.

1996-01-01

An operation for predicting a critical point by using a existent reverse multiplication method has been complicated, and an effective multiplication factor could not be plotted directly to degrade the accuracy for the prediction. The present invention comprises a detector counting memory section for memorizing the counting sent from a power detector which monitors the reactor power, a reverse multiplication factor calculation section for calculating the reverse multiplication factor based on initial countings and current countings of the power detector, and a critical point prediction section for predicting the criticality by the reverse multiplication method relative to effective multiplication factors corresponding to the state of the reactor core previously determined depending on the cases. In addition, a reactor core characteristic calculation section is added for analyzing an effective multiplication factor depending on the state of the reactor core. Then, if the margin up to the criticality is reduced to lower than a predetermined value during critical operation, an alarm is generated to stop the critical operation when generation of a period of more than a predetermined value predicted by succeeding critical operation. With such procedures, forecasting for the critical point can be easily predicted upon critical operation to greatly mitigate an operator's burden and improve handling for the operation. (N.H.)

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

Data acquisition for the LVR-15 research reactor. Final report

International Nuclear Information System (INIS)

Dusek, J.; Holy, J.; Rysavy, J.

1993-11-01

The activities are reviewed carried out under contract No. 5686 between the IAEA and the Nuclear Research Institute at Rez. A list of components, their description and block diagrams of the LVR-15 reactor are presented. Totally, 40 failures during testing and 48 failures during operation were recorded for the period 1991 to 1993. The failure causes and development are briefly described. Information on the failures was classified and included into the system. The contribution to data classification and processing is presented. A number of additional variants are pointed out, of the exact use of non-parametric and parametric statistical methods when developing the comprehensive probabilistic model. A list is given of initiating events as starting points of accident sequences collected from the operating experience. The report consists of three supplements: (i) Data collection on the LVR-15 research reactor; (ii) Some statistical methods for the data processing; (iii) Initiating events data of research reactor for the use of probabilistic safety assessment. (J.B.) 54 tabs., 17 figs., 14 refs

New reactor concepts

International Nuclear Information System (INIS)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

The role of point defect clusters in reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Stoller, R.E.

1993-01-01

Radiation-induced point defect clusters (PDC) are a plausible source of matrix hardening in reactor pressure vessel (RPV) steels in addition to copper-rich precipitates. These PDCs can be of either interstitial or vacancy type, and could exist in either 2 or 3-D shapes, e.g. small loops, voids, or stacking fault tetrahedra. Formation and evolution of PDCs are primarily determined by displacement damage rate and irradiation temperature. There is experimental evidence that size distributions of these clusters are also influenced by impurities such as copper. A theoretical model has been developed to investigate potential role of PDCs in RPV embrittlement. The model includes a detailed description of interstitial cluster population; vacancy clusters are treated in a more approximate fashion. The model has been used to examine a broad range of irradiation and material parameters. Results indicate that magnitude of hardening increment due to these clusters can be comparable to that attributed to copper precipitates. Both interstitial and vacancy type defects contribute to this hardening, with their relative importance determined by the specific irradiation conditions

Nuclear technology review 2003 update

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-09-01

Worldwide there were 441 nuclear power plants (NPPs) operating at the end of 2002.These supplied 16% of global electricity generation in 2002, down slightly from 16.2% in 2001.1 Table 1 summarizes world nuclear experience as of the end of 2002. The global energy availability factor for NPPs rose to 83.4% in 2001, from 82.1% in 2000 and 74.2% in 1991. In 2002, upratings calculated from data on the IAEA's Power Reactor Information System (PRIS) totalled approximately 672 MW(e), of which the United States of America accounted for 574 MW(e) and the United Kingdom accounted for 98 MW(e).The United States Nuclear Regulatory Commission (NRC) expects applications for 2270 MW(e) worth of upratings over the next five years. Six new NPPs were connected to the grid in 2000, three in 2001, and six in 2002. There were three retirements in 2000: Chernobyl-3 in Ukraine and two units at Hinkley Point A in the United Kingdom.There were no retirements in 2001 and four in 2002:Kozloduy-1 and -2 in Bulgaria and Bradwell units A and B in the UK. In 2002, construction started on seven new NPPs: six in India and one in the Democratic People's Republic of Korea. This issue covers the following topics: Medium-Term Projections; Sustainable Development; Resources And Fuel; Decommissioning; Advanced Designs; Research Reactors; Waste From Non-Power Applications; Nuclear Knowledge; Matters Of Interest To The IAEA Arising From The World Summit On Sustainable Development; International Project On Innovative Nuclear Reactors And Fuel Cycles (INPRO); Knowledge Management; Key Commitments, Targets And Timetables From The Johannesburg Plan Of Implementation; Management Of The Natural Resource Base.

Nuclear technology review 2003 update

International Nuclear Information System (INIS)

2003-09-01

Worldwide there were 441 nuclear power plants (NPPs) operating at the end of 2002.These supplied 16% of global electricity generation in 2002, down slightly from 16.2% in 2001.1 Table 1 summarizes world nuclear experience as of the end of 2002. The global energy availability factor for NPPs rose to 83.4% in 2001, from 82.1% in 2000 and 74.2% in 1991. In 2002, upratings calculated from data on the IAEA's Power Reactor Information System (PRIS) totalled approximately 672 MW(e), of which the United States of America accounted for 574 MW(e) and the United Kingdom accounted for 98 MW(e).The United States Nuclear Regulatory Commission (NRC) expects applications for 2270 MW(e) worth of upratings over the next five years. Six new NPPs were connected to the grid in 2000, three in 2001, and six in 2002. There were three retirements in 2000: Chernobyl-3 in Ukraine and two units at Hinkley Point A in the United Kingdom.There were no retirements in 2001 and four in 2002:Kozloduy-1 and -2 in Bulgaria and Bradwell units A and B in the UK. In 2002, construction started on seven new NPPs: six in India and one in the Democratic People's Republic of Korea. This issue covers the following topics: Medium-Term Projections; Sustainable Development; Resources And Fuel; Decommissioning; Advanced Designs; Research Reactors; Waste From Non-Power Applications; Nuclear Knowledge; Matters Of Interest To The IAEA Arising From The World Summit On Sustainable Development; International Project On Innovative Nuclear Reactors And Fuel Cycles (INPRO); Knowledge Management; Key Commitments, Targets And Timetables From The Johannesburg Plan Of Implementation; Management Of The Natural Resource Base

Reactor safety protection system

International Nuclear Information System (INIS)

Nishi, Hiroshi; Yokoyama, Tsuguo.

1989-01-01

A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

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.

Reactor core cooling device

International Nuclear Information System (INIS)

Kobayashi, Masahiro.

1986-01-01

Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

International Nuclear Information System (INIS)

Abouhadra, D.S.; Byrne, J.E.

2004-01-01

In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

R-102, 1 Group Space-Independent Inverse Reactor Kinetics

International Nuclear Information System (INIS)

Kaganove, J.J.

1966-01-01

1 - Description of problem or function: Given the space-independent, one energy group reactor kinetics equations and the initial conditions, this program determines the time variation of reactivity required to produce the given input of flux-time data. 2 - Method of solution: Time derivatives of neutron density are obtained by application of (a) five-point quartic, (b) three-point parabolic, (c) five-point least-mean-square cubic, (d) five-point least-mean-square parabolic, or (e) five-point least-mean-square linear formulae to the neutron density or to the natural logarithm of the neutron density. Between each data point the neutron density is assumed to be (a) exponential*(third-order polynomial), (b) exponential, or (c) linear. Changes in reactivity between data points are obtained algebraically from the kinetics equations, neutron density derivatives, and the algebraic representation of neutron density. First and second time derivatives of the reactivity are obtained by use of any of the formulae applicable to the neutron density. 3 - Restrictions on the complexity of the problem: Maxima of - 50 delay groups; 1000 data points; 99 data blocks (A data block is a sequence of input points characterized by a fixed time-interval between points, a smoothing option, and a number of repetitions of the smoothing option)

Subcutaneous injection of sup 99m Tc pertechnetate at acupuncture points K-3 and B-60

Energy Technology Data Exchange (ETDEWEB)

Wu Chung-Chieng; Jong Shiang-Bin; Lin Chun-Ching; Chen Min-Fen; Chen Jong-Rern (Kaohsiung Medical Coll., Taiwan (China)); Chung Chieng

1990-06-01

The acupuncture points are known to be morphologically related to the nerves and vessels. Yet the physiological role of blood vessels in the formation of acupuncture points remains unknown. With subcutaneous injection of {sup 99m}Tc pertechnetate at the acupuncture points K-3 and B-60 and with intra-acupuncture point injection of {sup 99m}Tc pertechnetate at K-3 and B-60, a lower-limb venography like what was obtained by intravenous injection of {sup 99m}Tc macroaggregated albumin was demonstrated in the present study. It seems that some acupuncture points do play a role in drainage of tissue fluid from soft tissue into the veins. (author).

Cosmiclike domain walls in superfluid 3He-B: Instantons and diabolical points in (k,r) space

International Nuclear Information System (INIS)

Salomaa, M.M.; Volovik, G.E.

1988-01-01

The possible planar superfluid B-B boundaries between inequivalent B-phase vacua are considered; such B-B interfaces provide an analogy with the cosmic domain walls that are believed to have precipitated in the phase transitions of the early Universe. Several of them display nontrivial structure in (k,r) space (i.e., the union of the momentum and real spaces). Such a wall represents an instanton connecting two B-phase vacua with different k-space topology. The transition between the vacua occurs through the formation of a pointlike defect either in the (k,r) space, or in the (k,t) space. These defects are so-called diabolical points of codimension 4, at which the fermionic energy tends to zero, thus providing the fermionic zero modes. Such points are new examples (within condensed-matter physics) of the peculiar diabolical points, which are characterized by the occurrence of a contact between the different branches of the quasiparticle spectra; in the present case, the branches of particles and holes, respectively. These points are here discussed for the case of the superfluid phases of liquid 3 He in close analogy with the quantum field theory of fermions interacting with classical bosonic fields. The cosmiclike domain walls in superfluid 3 He-B are observable in principle; in particular, the motion of the superfluid A-B interface is governed at low temperatures by the periodical emission of these topological excitation planes

Choosing the Optimal Number of B-spline Control Points (Part 1: Methodology and Approximation of Curves)

Science.gov (United States)

Harmening, Corinna; Neuner, Hans

2016-09-01

Due to the establishment of terrestrial laser scanner, the analysis strategies in engineering geodesy change from pointwise approaches to areal ones. These areal analysis strategies are commonly built on the modelling of the acquired point clouds. Freeform curves and surfaces like B-spline curves/surfaces are one possible approach to obtain space continuous information. A variety of parameters determines the B-spline's appearance; the B-spline's complexity is mostly determined by the number of control points. Usually, this number of control points is chosen quite arbitrarily by intuitive trial-and-error-procedures. In this paper, the Akaike Information Criterion and the Bayesian Information Criterion are investigated with regard to a justified and reproducible choice of the optimal number of control points of B-spline curves. Additionally, we develop a method which is based on the structural risk minimization of the statistical learning theory. Unlike the Akaike and the Bayesian Information Criteria this method doesn't use the number of parameters as complexity measure of the approximating functions but their Vapnik-Chervonenkis-dimension. Furthermore, it is also valid for non-linear models. Thus, the three methods differ in their target function to be minimized and consequently in their definition of optimality. The present paper will be continued by a second paper dealing with the choice of the optimal number of control points of B-spline surfaces.

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

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

The RED (reduce everyone's dose) initiative at Hinkley Point Power Station

International Nuclear Information System (INIS)

Weston, J.

1995-01-01

This paper does not offer a universal solution to reducing doses nor does it claim to introduce radically new ideas. It is a summarised account of the successes achieved in reducing the collective dose at a 30 year old power station. The key areas identified are the rewards of teamwork, the need to challenge established practices and the benefit of a dosimetry database. (author)

The economic failure of nuclear power in Britain

International Nuclear Information System (INIS)

Henney, A.

1990-01-01

Claims made about the economics of nuclear power have been misleading. The history and political framework within which nuclear power has developed in Britain are explained so that those claims can be understood. The main factors affecting the development of nuclear power in Britain have been military requirements, national pride, the hope of cheap electricity and concern about the security of fuel supplies. Variations in the official view of the economics of Magnox reactors are used to illustrate changes in the government attitude to nuclear power economics. Other factors - the 'oil crisis' of the 1970's the miners' strike, the accident at Three Mile Island and methods of accounting are all shown to influence this attitude. At the Hinkley Point C Inquiry the Central Electricity Generating Board conceded that nuclear power was not economic a position recognised by the government in the non-privatisation of nuclear power. (UK)

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}.

Analysis of boiling water reactors capacities for the 100% MOX fuel recycling

International Nuclear Information System (INIS)

Knoche, Dietrich

1999-01-01

The electro-nuclear park exploitation leads to plutonium production. The plutonium recycling in boiling water reactors performs a use possibility. The difference between the neutronic characteristics of the uranium and the plutonium need to evaluate the substitution impact of UOX fuel by MOX fuel on the reactor operating and safety. The analysis of the main points reached to the following conclusions: the reactivity coefficients are negative, during a cooling accident the re-divergence depends on the isotopic vector of the used plutonium, the efficiency lost of control cross resulting from the plutonium utilization can be compensate by the increase of the B 4C enrichment by 10 B and the change of the steel structure by an hafnium structure, the reactivity control in evolution can be obtained by the fuel poisoning (gadolinium, erbium) and the power map control by the plutonium content monitoring. (A.L.B.)

A highly accurate benchmark for reactor point kinetics with feedback

International Nuclear Information System (INIS)

Ganapol, B. D.; Picca, P.

2010-10-01

This work apply the concept of convergence acceleration, also known as extrapolation, to find the solution to the reactor kinetics equations describing nuclear reactor transients. The method features simplicity in that an approximate finite difference formulation is constructed and converged to high accuracy from knowledge of how the error term behaves. Through Rom berg extrapolation, we demonstrate its high accuracy for a variety of imposed reactivity insertions found in the literature as well as nonlinear temperature and fission product feedback. A unique feature of the proposed method, called RKE/R(om berg) algorithm, is interval bisection to ensure high accuracy. (Author)

A CAREM reactor's design evaluation from the nuclear security point of view

International Nuclear Information System (INIS)

Kay, J.M.; Felizia, E.R.; Navarro, N.R.; Caruso, G.J.

1990-01-01

The main objective of this work is to define the adequate rules for CAREM reactor security systems design and processes which aim to assure verification of the CALIN regulations 'Radiological Criteria' in relation to accidents concerning CAREM reactor design. (Author) [es

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

Fort St. Vrain high temperature gas-cooled reactor. Pt. 12. The dew point moisture monitor testing program

Energy Technology Data Exchange (ETDEWEB)

Olson, H.G. (Colorado State Univ., Fort Collins (USA). Dept. of Mechanical Engineering); Brey, H.L. (Public Service Co. of Colorado, Denver (USA)); Swart, F.E. (Gas-Cooled Reactor Associates, La Jolla, CA (USA)); Forbis, J.M. (Storage Technology Corp., Louisville, CO (USA))

1982-09-01

Moisture ingress into the core volume could cause damaging reactions with the moderator-reflector graphite and burnable poison, therefore a dew point moisture monitoring system has been developed with the basic design criteria that a plant protective system trip is signaled after the system detects high primary coolant helium moisture levels and that the system is able to correctly identify which of two steam generator loops is leaking. Modifications to the sample supplies to the monitors were necessary to reduce the system's unsatisfactory response time at lower reactor power levels.

Point 2004 A Temperature Dependent ENDF/B-VI, Release 8 Cross Section Library

International Nuclear Information System (INIS)

Cullen, D E

2004-01-01

The ENDF/B data library has recently been updated and is now freely available through the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. This most recent library is identified as ENDF/B-VI, Release 8. Release 8 completely supersedes all preceding releases. Release 8 will be the last release of ENDF/B-VI; the next release of ENDF/B data will be for the new ENDF/B-VII library. As distributed the ENDF/B-VI, Release 8 data includes cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in applications this library has been processed into the form of temperature dependent cross sections at eight neutron reactor like temperatures, between 0 and 2100 Kelvin, in steps of 300 Kelvin. It has also been processed to five astrophysics like temperatures, 1, 10, 100 eV, 1 and 10 keV. For reference purposes, 300 Kelvin is approximately 1/40 eV, so that 1 eV is approximately 12,000 Kelvin. At each temperature the cross sections are tabulated and linearly interpolable in energy. All results are in the computer independent ENDF/B-VI character format [1], which allows the data to be easily transported between computers. In its processed form this library is approximately 4.3 gigabyte in size and is distributed on a single DVD

Oscillation characteristics of the reactor 'A'; Oscilatorne karakteristike reaktora 'A'

Energy Technology Data Exchange (ETDEWEB)

Zecevic, V; Lolic, B [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1961-07-01

In addition to good knowledge of reactor physical properties, design of the reactor oscillator demands determining of the oscillator operating points as well as oscillation reactor properties. This paper contains study of the RA reactor power changes due to oscillations in in one of the vertical experimental channels. It has been concluded that the reactor optimum operating conditions are attained when the oscillator operates at optimum points, and other parameters are determined dependent on the sensitivity of the method and reactor stability.

Dosimetry system of the RB reactor

International Nuclear Information System (INIS)

Lolic, B.; Vukadin, D.

1962-01-01

Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

Investigating the Impact of Asp181 Point Mutations on Interactions between PTP1B and Phosphotyrosine Substrate

Science.gov (United States)

Liu, Mengyuan; Wang, Lushan; Sun, Xun; Zhao, Xian

2014-05-01

Protein tyrosine phosphatase 1B (PTP1B) is a key negative regulator of insulin and leptin signaling, which suggests that it is an attractive therapeutic target in type II diabetes and obesity. The aim of this research is to explore residues which interact with phosphotyrosine substrate can be affected by D181 point mutations and lead to increased substrate binding. To achieve this goal, molecular dynamics simulations were performed on wild type (WT) and two mutated PTP1B/substrate complexes. The cross-correlation and principal component analyses show that point mutations can affect the motions of some residues in the active site of PTP1B. Moreover, the hydrogen bond and energy decomposition analyses indicate that apart from residue 181, point mutations have influence on the interactions of substrate with several residues in the active site of PTP1B.

Method of operating FBR type reactors

International Nuclear Information System (INIS)

Arie, Kazuo.

1984-01-01

Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

Mirror hybrid reactor studies

International Nuclear Information System (INIS)

Bender, D.J.

1978-01-01

The hybrid reactor studies are reviewed. The optimization of the point design and work on a reference design are described. The status of the nuclear analysis of fast spectrum blankets, systems studies for fissile fuel producing hybrid reactor, and the mechanical design of the machine are reviewed

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

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

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.

IAU 2015 Resolution B2 on Recommended Zero Points for the Absolute and Apparent Bolometric Magnitude Scales

DEFF Research Database (Denmark)

Mamajek, E. E.; Torres, G.; Prsa, A.

2015-01-01

The XXIXth IAU General Assembly in Honolulu adopted IAU 2015 Resolution B2 on recommended zero points for the absolute and apparent bolometric magnitude scales. The resolution was proposed by the IAU Inter-Division A-G Working Group on Nominal Units for Stellar and Planetary Astronomy after...... consulting with a broad spectrum of researchers from the astronomical community. Resolution B2 resolves the long-standing absence of an internationally-adopted zero point for the absolute and apparent bolometric magnitude scales. Resolution B2 defines the zero point of the absolute bolometric magnitude scale...... such that a radiation source with $M_{\\rm Bol}$ = 0 has luminosity L$_{\\circ}$ = 3.0128e28 W. The zero point of the apparent bolometric magnitude scale ($m_{\\rm Bol}$ = 0) corresponds to irradiance $f_{\\circ}$ = 2.518021002e-8 W/m$^2$. The zero points were chosen so that the nominal solar luminosity (3.828e26 W...

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

Science.gov (United States)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

International Nuclear Information System (INIS)

Shen Yongjun; Ding Jiandong; Lei Lecheng; Zhang Xingwang

2014-01-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10 −9 mol/L and 0.61 × 10 −9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10 −2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10 −2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation

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.)

Nuclear reactors theory

International Nuclear Information System (INIS)

Naudan, G.; Nigon, J.L.

1993-01-01

After principles of chain reaction and criticality notion, a descriptive model of neutrons behaviour is exposed from a local point of view (this model is called four factors model). One justifies the use of middle values for the calculation of the distribution in space of reactor, quantities representing heterogeneous middle from a local point of view (fuel, moderator, can or clad, and so on ...) by substitution of an equivalent homogeneous middle. Time dependence, dynamical behaviour of reactor are studied. Long term effects of evolution of constituents elements of heart under irradiation, and ways to balance this evolution are in the last paragraph. 18 refs., 26 figs

Licensing of the TRIGA Mark III reactor at the Mexican Nuclear Centre

International Nuclear Information System (INIS)

Ramirez, R.M.; Arrendondo, R.R.

1990-01-01

The TRIGA Mark III reactor at the Mexican Nuclear Centre went critical in 1968 and remained so until 1979 when the National Commission for Nuclear Safety and Safeguards (CNSNS), the Mexican regulatory authority, was set up. The reactor was therefore operating without a formal operating license, and the CNSNS accordingly requested the ININ to license the reactor under the existing conditions and to ensure that any modification of the original design complied with Standards ANSI/ANS-15 and with the code of practice set out in IAEA Safety Series No. 35. The most relevant points in granting the operating licence were: (a) the preparation of the Safety Report; (b) the formulation and application of the Quality Assurance Programme; (c) the reconditioning of the following reactor systems: the cooling systems; the ventilation and exhaust system; the monitoring system and control panel; (d) the training of the reactor operating staff at junior and senior levels; and (e) the formulation of procedures and instructions. Once the provisional operating license was obtained for the reactor it was considered necessary to modify the reactor core, which has been composed of 20% enriched standards fuel, to a mixed core based on a mixture of standard fuel and FLIP-type fuel with 70% 235 U enrichment. The CNSNS therefore requested that the mixed core be licensed and a technical report was accordingly annexed to the Safety Report, its contents including the following subjects: (a) neutron analysis of the proposed configuration; (b) reactor shutdown margins; (c) accident analysis; and (d) technical specifications. The licensing process was completed this year and we are now hoping to obtain the final operating license

Study of radiation exposure rate on the measurement points in Kartini reactor hall as based to determine operation safety parameters (KBO)

International Nuclear Information System (INIS)

Mahrus Salam; Elisabeth Supriyatni; Fajar Panuntun

2016-01-01

In the operation of nuclear facility there are safety parameters, which is the value of the conservatively maximum limit to ensure that all of the uncertainty in the analysis of facility operations safety have been considered, such as uncertainty of measurement, response time and uncertainty calculation tool, and is get a long to others value of normal operating condition limits, in other words, there are still allowed or permitted. Calculation of the radiation exposure rate on five measurement points (50 cm above the water surface of reactor pool, above interim storage (bulk shielding), reactor deck, thermal column and sub critical facility) and to be compared to the operation safety parameters (KBO) of Kartini reactor. The exposure rate value is obtained by calculating the source term of radioactivity on the core, attenuation resulting from the radiation shielding and measurement distance. From the calculation obtained that the value of gamma exposure rate of 50 cm above the water surface of reactor pool is 96.91 mR/hr (KBO<100 mR/hr), on the deck of Bulk Shielding amounted to 1.70 mR/h (KBO<2.5 mR/hr), on the reactor deck amounted to 5.73 mR/hr (KBO<10 mR/hr), on the Thermal Column amounted to 2.73 mR/hr (KBO<10 mR/hr) and on the sub critical facility amounted to 1.148 mR/hr (KBO<2.5 mR/hr). The value of gamma exposure rate at 5 locations measurements are still less than the operation safety parameters (KBO), it means that the reactor is safe to be operated. (author)

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

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

The analysis of thermal-hydraulic performances of nuclear ship reactor

International Nuclear Information System (INIS)

Wakabayashi, Shinshichi; Hamada, Masao

1975-01-01

Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

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

Areva: 1. quarter 2015 revenue down, at euros 1.762 bn: -1.1% vs. March 2014 (-0.9% like for like)

International Nuclear Information System (INIS)

Repaire, Philippine du

2015-01-01

In the 1. quarter of 2015, AREVA generated consolidated revenue of 1.762 billion euros, representing a decrease of 1.1% (-0.9% like for like) compared with the same period in 2014. Foreign exchange had a positive impact of 36 million euros over the period, while consolidation scope had a negative impact of 39 million euros. At March 31, 2015, the group had 47.520 billion euros in backlog, a 1.4% increase in relation to December 31, 2014 (46.866 billion euros) reflecting a favorable foreign exchange impact. It should be noted that the backlog does not include the amount from agreements signed with EDF in October 2013 for the EPR reactors project at Hinkley Point in the United Kingdom or for the related fuel. The order intake totaled 881 million euros in the 1. quarter of 2015, an increase compared with the 1. quarter of 2014 (668 million euros)

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.

Reactor for exothermic reactions

Science.gov (United States)

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

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

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (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.

Nuclear reactor, fuel assembly and neutron measuring system

International Nuclear Information System (INIS)

Chaki, Masao; Murase, Michio; Zukeran, Atsushi; Moriya, Kimiaki

1998-01-01

The present invention provides a BWR type reactor improved with the efficiency of used fuels and fuel economy by increasing a rated power and reducing exchange fuels. Namely, in a BWR type reactor at present, a thermal limit value is determined by conducting nuclear calculation of the reactor core based on data of reactor flow rate measurement and data of neutron flux measurement. However, since the neutron calculation of the reactor core is based on fuel assemblies while the points for the neutron measurement are present at the outside of the fuel assemblies, errors are caused. A margin including the errors has been used as a thermal limit value during operation. In the present invention, neutron fluxes in the fuel assembly as a base of the nuclear calculation can be measured by the same number of neutron detector tubes, but the number of the measuring points is increased to four times. With such procedures, errors caused by the difference of the neutron calculation and values at neutron measuring points can be reduced. As a result, a margin of the thermal limit value is reduced to increase the degree of freedom of reactor operation. Then, the economical property of the reactor operation can be improved. (N.H.)

Status of Japanese university reactors

International Nuclear Information System (INIS)

Fujita, Yoshiaki

1999-01-01

Status of Japanese university reactors, their role and value in research and education, and the spent fuel problem are presented. Some of the reactors are now faced by severe difficulties in continuing their operation services. The point of measures to solve the difficulties is suggested. (author)

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

Reactor shutdown device

International Nuclear Information System (INIS)

Ito, Masahiko

1990-01-01

The object of the present invention is to reliably shutdown an LMFBR type reactor upon accident of the reactor. That is, curie point magnetic member is made annular so that it can be moved between the outer circumference of an electromagnet and the position above the electromagnet. This enables to enlarge the curie point magnetic member since it is no more necessary to be inserted it in a guide tube. Accordingly, attracting force upon normal operation is increased to remarkably improve the reliability against erronerous scram, etc. Further, since a required gap is formed between the curie point magnetic member and the electromagnet and the heat of coolants is efficiently transmitted to the curie point magnetic member, rapid scram is possible. Further, a position support mechanism is disposed to a part of a control element or at the inner side of the guiding tube for urging and actuating the armature to make it protrude above the top of the guiding tube. With such a constitution, since the armature can be adsorbed without inserting the curie point magnetic member and the electromagnet guide tube, the same effect as in the case of inserting them can be obtained. (I.S.)

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

Science.gov (United States)

El-Genk, Mohamed S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

International Nuclear Information System (INIS)

El-genk, M.S.

1991-01-01

The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

Nucleaire et Energies Nr 65 - March 2015

International Nuclear Information System (INIS)

Lenail, Bernard; Ducroux, Guy; Seyve, Claude; Simonnet, Jacques; Justin, Francois; Salanave, Jean-Luc; Raisonnier, Daniele

2015-03-01

After a first brief article about the difficult situation faced by Areva, and an article about the Japanese situation four years after the Fukushima accident and the progressive restarting of reactors, an article proposes an overview of recent evolutions and events in the energy sector: the sharp decrease of oil prices and its consequences for Total and the European Union. Three articles concern nuclear activities: recent events in the mining and fuel fabrication sectors (perspectives, restructuring in the USA, contracts for Areva and EDF), overview of the situation of reactors (with perspective for the installed power by 2050, unclear perspectives in France, projects in the UK in Hinkley Point, Moorside, and Wylfa, situation in Belgium, Hungary and Russia, contracts for Areva and EDF, situation in Taiwan, India, South Korea, Japan, USA, South Africa), overview of activities and events related to the back-end of the fuel cycle and to decommissioning in different countries (recycling activities in France, decommissioning in France and in the UK, warehousing in France and Germany, a new warehousing system to be approved in the USA, events related to storage activities). An article discusses issues related to the lifetime of French nuclear reactors, and another one discusses the identification of electric power costs and ways to compare them. A last article briefly discusses some terminological confusion between energy and electricity, and between produced energy and available power

Determination of the decay constants and relative abundances of delayed neutrons by noise analysis in zero-power reactors

International Nuclear Information System (INIS)

Diniz, Ricardo

2005-01-01

A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters β i and λ i in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the β eff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)

Welding in repair of nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Pilous, V.; Kovarik, R.

1987-01-01

Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

Supplementary control points for reactor shutdown without access to the main control room (International Electrotechnical Commission Standard Publication 965:1989)

International Nuclear Information System (INIS)

Kubalek, J.; Hajek, B.

1993-01-01

This standard establishes the requirements for supplementary Control Points provided to enable the operating staff to shut down the reactor and maintain the plant in a safe shut-down condition when the main control room is no longer available. This standard covers the functional selection, design and organization of the man/machine interface. It also establishes requirements for procedures which systematically verify and validate the functional design of supplementary control points. The requirements reflect the application of human engineering principles as they apply to man/machine interface. This standard does not cover special emergency response centres (e.g. a Technical Support Centre). It also does not include the detailed equipment design. Unavailability of the main control room controls due to intentionally man-induced events is not considered

Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

Energy Technology Data Exchange (ETDEWEB)

Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-09-01

An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

The fast breeder reactor

International Nuclear Information System (INIS)

Davis, D.A.; Baker, M.A.W.; Hall, R.S.

1990-01-01

Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

Computation of point reactor dynamics equations with thermal feedback via weighted residue method

International Nuclear Information System (INIS)

Suo Changan; Liu Xiaoming

1986-01-01

Point reactor dynamics equations with six groups of delayed neutrons have been computed via weighted-residual method in which the delta function was taken as a weighting function, and the parabolic with or without exponential factor as a trial function respectively for an insertion of large or smaller reactivity. The reactivity inserted into core can be varied with time, including insertion in forms of step function, polynomials up to second power and sine function. A thermal feedback of single flow channel model was added in. The thermal equations concerned were treated by use of a backward difference technique. A WRK code has been worked out, including implementation of an automatic selection of time span based on an input of error requirement and of an automatic change between computation with large reactivity and that with smaller one. On the condition of power varied slowly and without feedback, the results are not sensitive to the selection of values of time span. At last, the comparison of relevant results has shown that the agreement is quite well

Characteristics of a reactor with power reactivity feedback

International Nuclear Information System (INIS)

Li Fengyu; Zhang Yusheng; Zhang Guangfu; Liu Ying

2008-01-01

The point-reactor model with power reactivity feedback becomes a nonlinear system. Its dynamic characteristic shows great complexity. According to the mathematic definition of stability in differential equation qualitative theory, the model of a reactor with power reactivity feedback is judged unstable. The equilibrium point is a saddle-node point. A portion of the trajectory in the neighborhood of the equilibrium point is parabolic fan curve, and the other is hyperbolic fan curve. Based on phase locus near the equilibrium point, it is pointed out that the model is still stable within physical limits. The difference between stabilities in the mathematical sense and in the physical sense is indicated. (authors)

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)

Neutron density fluctuations in point reactor systems with dichotomic reactivity noise

International Nuclear Information System (INIS)

Sako, Okitsugu

1984-01-01

The exactly solvable stochastic point reactor model systems are analyzed through the stochastic Liouville equation. Three kinds of model systems are treated: (1) linear system without delayed neutrons, (2) linear system with one-group of delayed neutrons, and (3) nonlinear system with direct power feedback. The exact expressions for the fluctuations of neutron density, such as the moments, autocorrelation function and power spectral density, are derived in the case where the colored reactivity noise is described by the dichotomic, or two state, Markov process with arbitrary correlation time and intensity, and the effects of the finite correlation time and intensity of the noise on the neutron density fluctuations are investigated. The influence of presence of delayed neutrons and the effect of nonlinearity of system on the neutron density fluctuations are also elucidated. When the reactivity correlation time is very short, the correlation time has almost no effect on the power spectral density, and the relative fluctuation of neutron density in the stationary state is not affected very much by the presence of delayed neutrons and also by the nonlinearity of system. On the other hand, if the reactivity correlation time is very long, the effect of the reactivity noise on the power spectral density appears at very low frequency, and the presence of delayed neutrons has an effect of reducing the neutron density fluctuations. (author)

Towards Compact Antineutrino Detectors for Safeguarding Nuclear Reactors

International Nuclear Information System (INIS)

Meijer, R.J. de; Smit, F.D.; Woertche, H.J.

2010-01-01

In 2008 the IAEA Division of Technical Support convened a Workshop on Antineutrino Detection for Safeguards Applications. Two of the recommendations expressed that IAEA should consider antineutrino detection and monitoring in its current R and D program for safeguarding bulk-process reactors, and consider antineutrino detection and monitoring in its Safeguards by Design approaches for power and fissile inventory monitoring of new and next generation reactors. The workshop came to these recommendations after having assessed the results obtained at the San Onofre Nuclear Generator Station (SONGS) in California. A 600 litre, 10% efficiency detector, placed at 25m from the core was shown to record 300 net antineutrino events per day. The 2*2.5*2.5 m 3 footprint of the detector and the required below background operation, prevents an easy deployment at reactors. Moreover it does not provide spatial information of the fissile inventory and, because of the shape of a PBMR reactor, would not be representative for such type of reactor. A solution to this drawback is to develop more efficient detectors that are less bulky and less sensitive to cosmic and natural radiation backgrounds. Antineutrino detection in the SONGS detector is based on the capture of antineutrinos by a proton resulting in a positron and neutron. In the SONGS detector the positron and neutron are detected by secondary gamma-rays. The efficiency of the SONGS detector is largely dominated by the low efficiency for gamma detection high background sensitivity We are investigating two methods to resolve this problem, both leading to more compact detectors, which in a modular set up also will provide spatial information. One is based on detecting the positrons on their slowdown signal and the neutrons by capturing in 10 B or 6 Li, resulting in alpha-emission. The drawback for standard liquid scintillators doped with e.g. B is the low flame point of the solvent and the strong quenching of the alpha signal. Our

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)

Qualification of the Taiwan Power Company's pressurized water reactor physics methods using CASMO-4/SIMULATE-3

Energy Technology Data Exchange (ETDEWEB)

Lin, Hsien-Chuan, E-mail: linsc@iner.org.tw [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yaur, Shung-Jung; Lin, Tzung-Yi; Kuo, Weng-Sheng; Shiue, Jin-Yih; Huang, Yu-Lung [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Studsvik's core management system (CMS) was applied to Taiwan Power Company's pressurized water reactor. Black-Right-Pointing-Pointer Advanced calculation model of shutdown cooling, B-10 depletion and integrated pin exposure were introduced. Black-Right-Pointing-Pointer Core characteristic parameters such as boron letdown, low power physics test (LPPT) predictions, and reaction rate were validated to measurement data. Black-Right-Pointing-Pointer The uncertainty of each item was quantified. - Abstract: This paper presents the validation of Studsvik core management system (CMS) for application to the Maanshan units 1 and 2 reactor core physics analysis (Huang and Yang, 1994). The methodology was validated by demonstrating the ability to obtain accurate and reliable results for various conditions and applications. Core characteristic parameters such as boron letdown, low power physics test (LPPT) predictions, and reaction rate were validated. Analytical results have been compared to measured data and reliability factors of the method have been quantified.

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

Fast breeder reactor safety : a perspective

International Nuclear Information System (INIS)

Kale, R.D.

1992-01-01

Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with 239 Pu/ 238 U (unused or depleted) produces (breeds) more fissionable fuel material 239 Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert 232 Th into 233 U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the high chemical energy potential of sodium. These two issues are analysed and it is pointed that they are manageable by current design, construction and operational practices. Main findings of safety research during the last six to eight years in West European Countries and United States of America (US) are summarised. Three stage engineered safety provision incorporated into the design of the sodium cooled Fast Breeder Test Reactor (FBTR) commissioned at Kalpakkam are explained. The important design safety features of FBTR such as primary system containment, emergency core cooling, plant protection system, inherent safety features achieved through reactivity coefficients, and natural convection cooling are discussed. Theoretical analysis and experimental research in fast reactor safety carried out at the Indira Gandhi Centre for Atomic Research during the past some years are reviewed. (M.G.B.)

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

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.)

Mirror Advanced Reactor Study (MARS)

International Nuclear Information System (INIS)

Logan, B.G.

1983-01-01

Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made

The relationship between milligram-hours quantity and doses at points A and B in gynecologic brachytherapy

International Nuclear Information System (INIS)

Martins, H.L.; Albuquerque, L.F.

1987-01-01

A comparison of mghr prescrition (Fletcher's system) to doses at points A and B (Manchester's system) in gynecologic brachytherapy is presented. The dose rate at points A and B from computerized isodose curves is studied. The f ratio = (dose rates)/(mgRaeq) is measured by: the ovoid charge, the curvature of uterine probe, the shifting charge of uterine probes, the lateral shifting of extern orifice in relationship of extern orifice in relationship to the pelve, the separation between the ovoids and the angulation of uterine probe. (M.A.C.) [pt

Study on transient of fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Streck, E.E.

1988-01-01

The point kinetic equations for a Fluidized-Bed Nuclear Reactor are solved by the method of Hansen. Due to the time varying nature of the reactor volume, the equations have a non-conventional formulation (moving boundary problem), but the method of solution preserves its asymptotic convergence and efficiency characteristics under this formulation. A one dimensional and linearized thermal hydraulics feedback model was coupled to the point kinetic equations in order to obtain a more realistic representation of the reactor power. The resulting equations are solved by the Euler explicit method. (author)

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)

Introduction to magnetic fusion reactor design

International Nuclear Information System (INIS)

Watanabe, Kenji

1988-01-01

Trend of the tokamak reactor design works so far carried out is reviewed, and method of conceptual design for commercial fusion reactor is critically considered concerning the black-box conpepts. System-framework of the engineering of magnetic fusion (commercial) reactor design is proposed as four steps. Based on it the next design studies are recommended in parallel approaches for making real-overcome of reactor material problem, from the view point of technological realization and not from the economical one. Real trials are involved. (author)

Nuclear reactor in deep water

International Nuclear Information System (INIS)

Anon.

1980-01-01

Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

Optimization and control of a continuous polymerization reactor

Directory of Open Access Journals (Sweden)

L. A. Alvarez

2012-12-01

Full Text Available This work studies the optimization and control of a styrene polymerization reactor. The proposed strategy deals with the case where, because of market conditions and equipment deterioration, the optimal operating point of the continuous reactor is modified significantly along the operation time and the control system has to search for this optimum point, besides keeping the reactor system stable at any possible point. The approach considered here consists of three layers: the Real Time Optimization (RTO, the Model Predictive Control (MPC and a Target Calculation (TC that coordinates the communication between the two other layers and guarantees the stability of the whole structure. The proposed algorithm is simulated with the phenomenological model of a styrene polymerization reactor, which has been widely used as a benchmark for process control. The complete optimization structure for the styrene process including disturbances rejection is developed. The simulation results show the robustness of the proposed strategy and the capability to deal with disturbances while the economic objective is optimized.

Reevaluation of steam generator level trip set point

Energy Technology Data Exchange (ETDEWEB)

Shim, Yoon Sub; Soh, Dong Sub; Kim, Sung Oh; Jung, Se Won; Sung, Kang Sik; Lee, Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

The reactor trip by the low level of steam generator water accounts for a substantial portion of reactor scrams in a nuclear plant and the feasibility of modification of the steam generator water level trip system of YGN 1/2 was evaluated in this study. The study revealed removal of the reactor trip function from the SG water level trip system is not possible because of plant safety but relaxation of the trip set point by 9 % is feasible. The set point relaxation requires drilling of new holes for level measurement to operating steam generators. Characteristics of negative neutron flux rate trip and reactor trip were also reviewed as an additional work. Since the purpose of the trip system modification for reduction of a reactor scram frequency is not to satisfy legal requirements but to improve plant performance and the modification yields positive and negative aspects, the decision of actual modification needs to be made based on the results of this study and also the policy of a plant owner. 37 figs, 6 tabs, 14 refs. (Author).

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)

Super critical water reactors

International Nuclear Information System (INIS)

Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

2005-01-01

Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

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

Application of the Wiener-Hermite functional method to point reactor kinetics driven by random reactivity fluctuations

International Nuclear Information System (INIS)

Behringer, K.; Pineyro, J.; Mennig, J.

1990-06-01

The Wiener-Hermite functional (WHF) method has been applied to the point reactor kinetic equation excited by Gaussian random reactivity noise under stationary conditions. Delayed neutrons and any feedback effects are disregarded. The neutron steady-state value and the power spectral density (PSD) of the neutron flux have been calculated in a second order (WHF-2) approximation. Two cases are considered: in the first case, the noise source is low-pass white noise. In both cases the WHF-2 approximation of the neutron PSDs leads to relatively simple analytical expressions. The accuracy of the approach is determined by comparison with exact solutions of the problem. The investigations show that the WHF method is a powerful approximative tool for studying the nonlinear effects in the stochastic differential equation. (author) 5 figs., 29 refs

Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

International Nuclear Information System (INIS)

Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

1974-01-01

In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

Neutronic analysis of two-fluid thorium molten salt reactor

International Nuclear Information System (INIS)

Frybort, Jan; Vocka, Radim

2009-01-01

The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)

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.)

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

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.)

The Dutch debate on reactor safety

International Nuclear Information System (INIS)

Aldenkamp, F.; Biesiot, W.; Geerts, H.M.; Nienhuys, K.; Soppe, W.

1986-06-01

A survey is presented of the discussion on location sites for nuclear power plants in the Netherlands and the USA. It consists of two parts (A and B). This part (A) presents a summary of the discussion in the Netherlands. The name and contents of the reports which have been published are indicated briefly. Many of the Dutch reports refer to American studies on reactor safety. In order to obtain a good survey of these a description is presented of the development of the discussion in the USA after the publication of the well-known reactor safety study WASH-1400. After this survey of five important reports, an introduction is presented dealing with the background of the authors, the purposes, the main conclusions and the most salient points of criticism formulated by others. Under discussion are the reports of the IDCOR (nuclear power industry), ANS (American Nuclear Society), APS (American Physical Society), NRC (Nuclear Regulatory Commission) and UCS (Union of Concerned Scientists). 38 refs.; 5 figs.; 5 tabs

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

Inertial fusion reactor designs

International Nuclear Information System (INIS)

Meier, W.

1987-01-01

In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

ZZ POINT-2007, linearly interpolable ENDF/B-VII.0 data for 14 temperatures

International Nuclear Information System (INIS)

Cullen, Dermott E.

2007-01-01

A - Description or function: The ENDF/B data library, ENDF/B-VII.0 was processed into the form of temperature dependent cross sections. The original evaluated data include cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in applications, these ENDF/B-VII.0 data were processed into the form of temperature dependent cross sections at eight temperatures: 0, 300, 600, 900, 1200, 1500, 1800 and 2100 Kelvin. It has also been processed to six astrophysics like temperatures: 0.1, 1, 10, 100 eV, 1 and 10 keV. At each temperature the cross sections are tabulated and linearly interpolable in energy with a tolerance of 0.1 %. POINT 2007 contains all of the evaluations in the ENDF/B-VII general purpose library, which contains 78 new evaluations + 315 old ones: total 393 nuclides. It also includes 16 new elemental evaluations replaced by isotopic evaluations + 19 old ones. No special purpose ENDF/B-VII libraries, such as fission products, thermal scattering, photon interaction data are included. These evaluations include all cross sections over the energy range 10 e-5 eV to at least 20 MeV. The list of nuclides is indicated. B - Methods: The PREPRO 2007 code system was used to process the ENDF/B data. Listed below are the steps, including the PREPRO2007 codes, which were used to process the data in the order in which the codes were run. 1) Linearly interpolable, tabulated cross sections (LINEAR) 2) Including the resonance contribution (RECENT) 3) Doppler broaden all cross sections to temperature (SIGMA1) 4) Check data, define redundant cross sections by summation (FIXUP) 5) Update evaluation dictionary in MF/MT=1/451 (DICTIN) C - Restrictions: Due to recent changes in ENDF-6 Formats and Procedures only the latest version of the ENDF/B Pre-processing codes, namely PREPRO 2007, can be used to accurately process all current ENDF/B-VII evaluations. The use of

Occupational analysis for the Angra-1 reactor

International Nuclear Information System (INIS)

Moraes, A.

1991-01-01

Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

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.)

A systematic analysis of the Braitenberg vehicle 2b for point-like stimulus sources

International Nuclear Information System (INIS)

Rañó, Iñaki

2012-01-01

Braitenberg vehicles have been used experimentally for decades in robotics with limited empirical understanding. This paper presents the first mathematical model of the vehicle 2b, displaying so-called aggression behaviour, and analyses the possible trajectories for point-like smooth stimulus sources. This sensory-motor steering control mechanism is used to implement biologically grounded target approach, target-seeking or obstacle-avoidance behaviour. However, the analysis of the resulting model reveals that complex and unexpected trajectories can result even for point-like stimuli. We also prove how the implementation of the controller and the vehicle morphology interact to affect the behaviour of the vehicle. This work provides a better understanding of Braitenberg vehicle 2b, explains experimental results and paves the way for a formally grounded application on robotics as well as for a new way of understanding target seeking in biology. (paper)

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.)

Directions for improved fusion reactors

International Nuclear Information System (INIS)

Krakowski, R.A.; Miller, R.L.; Delene, J.G.

1986-01-01

Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts

Nuclear power: the future reassessed

Energy Technology Data Exchange (ETDEWEB)

Roberts, L [East Anglia Univ., Norwich (UK). Environmental Risk Assessment Unit (ERAU)

1991-02-01

In recommending that consent be given for the construction of a further Pressurized Water Reactor at Hinkley Point in Somerset, UK, the Inspector at the Public Inquiry underlined two major benefits: (i) the contribution an additional large nuclear plant would make to the strategic objective of diversity of supply, and (ii) the environmental benefits of nuclear power compared to many alternative forms of electricity generation. The major environmental advantages of nuclear power over fossil fuel combustion arise both because of the small amounts of fuel required - 1/18,000 compared to coal - thus minimizing transport needs and land use, and because of the virtual absence of atmospheric emissions from nuclear stations. Nuclear reactors emit no acid gases and the nuclear fuel cycle gives rise to only small amounts of carbon dioxide. An expansion of the nuclear option is often opposed on three grounds; the need to dispose of radioactive waste; the danger of the proliferation of nuclear weapons and the risk of a large scale accident. However all these doubts can be answered and the arguments supporting nuclear safety are summarized. It is argued that the contribution to primary energy demand in Europe could be doubled or trebled by 2020 with considerable benefits in overall safety environmental impacts at no extra cost. (author).

Nuclear power: the future reassessed

International Nuclear Information System (INIS)

Roberts, L.

1991-01-01

In recommending that consent be given for the construction of a further Pressurized Water Reactor at Hinkley Point in Somerset, UK, the Inspector at the Public Inquiry underlined two major benefits: (i) the contribution an additional large nuclear plant would make to the strategic objective of diversity of supply, and (ii) the environmental benefits of nuclear power compared to many alternative forms of electricity generation. The major environmental advantages of nuclear power over fossil fuel combustion arise both because of the small amounts of fuel required - 1/18,000 compared to coal - thus minimizing transport needs and land use, and because of the virtual absence of atmospheric emissions from nuclear stations. Nuclear reactors emit no acid gases and the nuclear fuel cycle gives rise to only small amounts of carbon dioxide. An expansion of the nuclear option is often opposed on three grounds; the need to dispose of radioactive waste; the danger of the proliferation of nuclear weapons and the risk of a large scale accident. However all these doubts can be answered and the arguments supporting nuclear safety are summarized. It is argued that the contribution to primary energy demand in Europe could be doubled or trebled by 2020 with considerable benefits in overall safety environmental impacts at no extra cost. (author)

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

Operational methods of the fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Borges, V.; Sefidvash, F.

1993-01-01

The operational curve of reactivity as a function of porosity of the Fluidized Bed Nuclear Reactor is presented. The strategies for start-up, shut-down and maintaining the reactor critical during operation are described. The inherent safety of the reactor from neutronic point of view under steady state condition is demonstrated. (author)

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.)

Detection of SBLOCA in the reactor of PHT system of Indian PHWR using GLR method

Energy Technology Data Exchange (ETDEWEB)

Chakrabarti, Dipankar [Indian Institute of Technology, Kanpur (India). Nuclear Engineering and Technology Programme

1990-01-01

Detection of Small Break Loss of Coolant Accident (SBLOCA) in nuclear power plants is important from the point of view of safety. Generalised Likelihood Ratio (GLR) test is one of the ways to detect faults like leak, controller bias etc. It can differentiate and diagnose different types of faults. A simplified state-space variable model of a PHWR reactor is developed and the utility of GLR method is investigated to detect leaks in the coolant channel in the reactor portion of the primary heat transport (PHT) system. A simple digital control system to control the outlet pressure of the reactor by manipulating the flow rate through the reactor is also developed. The results indicate that a leak of magnitude as low as 0.25% of the total flow rate through one coolant channel can be detected efficiently and promptly by this method. For instance a leak was detected within 3 minutes properly for 97 times out of 100 leaks simulated. (M.G.B.). 20 refs., 1 appendix.

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

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

Design and construction of small power reactors

International Nuclear Information System (INIS)

Tachi, Yasuo

1992-01-01

Small size reactors are considered to have many advantages over large-sized reactors. But at the same time, small size reactors show eventual disadvantages in economy. In this paper one of the possibilities to improve its basic disadvantage will be discussed from a manufacturer's point of view. The stress will be placed on the possibility and possible effects of adoption of Computer Aided Engineering. (author). 2 figs

Method for operating nuclear reactor

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

1978-01-01

Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

Experimental Breeder Reactor I Preservation Plan

Energy Technology Data Exchange (ETDEWEB)

Julie Braun

2006-10-01

Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

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

2013-01-01

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

Effects of extrinsic point defects in phosphorene: B, C, N, O and F Adatoms

OpenAIRE

Wang, Gaoxue; Pandey, Ravindra; Karna, Shashi P.

2015-01-01

Phosphorene is emerging as a promising 2D semiconducting material with a direct band gap and high carrier mobility. In this paper, we examine the role of the extrinsic point defects including surface adatoms in modifying the electronic properties of phosphorene using density functional theory. The surface adatoms considered are B, C, N, O and F with a [He] core electronic configuration. Our calculations show that B and C, with electronegativity close to P, prefer to break the sp3 bonds of pho...

Minor actinide transmutation using minor actinide burner reactors

International Nuclear Information System (INIS)

Mukaiyama, T.; Yoshida, H.; Gunji, Y.

1991-01-01

The concept of minor actinide burner reactor is proposed as an efficient way to transmute long-lived minor actinides in order to ease the burden of high-level radioactive waste disposal problem. Conceptual design study of minor actinide burner reactors was performed to obtain a reactor model with very hard neutron spectrum and very high neutron flux in which minor actinides can be fissioned efficiently. Two models of burner reactors were obtained, one with metal fuel core and the other with particle fuel core. Minor actinide transmutation by the actinide burner reactors is compared with that by power reactors from both the reactor physics and fuel cycle facilities view point. (author)

RCC-E: Design and construction rules for electrical equipment of PWR nuclear islands

International Nuclear Information System (INIS)

2016-01-01

RCC-E describes the rules for designing, building and installing electrical and I and C systems and equipment for pressurized water reactors. The code was drafted in partnership with industry, engineering firms, manufacturers, building control firms and operators, and represents a collection of best practices in accordance with IAEA requirements and IEC standards. The code's scope covers: architecture and the associated systems, materials engineering and the qualification procedure for normal and accidental environmental conditions, facility engineering and management of common cause failures (electrical and I and C) and electromagnetic interference, testing and inspecting electrical characteristics, quality assurance requirements supplementing ISO 9001 and activity monitoring. Use: RCC-E has been used to build the following power plants: France's last 12 nuclear units (1,300 MWe (8) and 1,450 MWe (4)), 2 M310 reactors in Korea (2), 44 M310 (4), CPR-1000 (28), CPR-600 (6), HPR-1000 (4) and EPR (2) reactors in service or undergoing construction in China, 1 EPR reactor in France. RCC-E is used for maintenance operations in French power plants (58 units) and Chinese M310 and CPR-1000 power plants. RCC-E has been chosen for the construction of the EPR plants in Hinkley Point, UK. Contents of the 2016 edition of the RCC-E code: Volume 1 - General requirements and quality assurance; Volume 2 - Specification of requirements; Volume 3 - I and C systems; Volume 4 - Electrical systems; Volume 5 - Materials engineering; Volume 6 - Installation of electrical and I and C systems; Volume 7 - Inspection and test methods

Development of self-actuated shutdown system using curie point electromagnet

International Nuclear Information System (INIS)

Kim, Tae Ryong; Park, Jin Ho

1999-01-01

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system (SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet (CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid Metal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design. (author)

Next Generation Reactors in Korea

International Nuclear Information System (INIS)

Oh, Yongshick; Choi, Youngsang; Park, Keecheol

1990-01-01

In Korea, nuclear power will be continuously needed to meet the trend of steady increase in electricity demand. But in relation to the further development of nuclear energy, there are still many uncertainties to be solved such as power demand forecast, site availability, thermal energy utilization and technology enhancement for economic and safety. To cope with those uncertainties effectively and to proceed the nuclear projects uninterruptedly, KEPCO decided to initiate two research project. i. e., one is 'the outlook and developmental strategy of nuclear energy for the early 21st century in the R. O. K' and the other is 'the feasibility study on the advanced reactors in Korea. Prospects of nuclear energy in Korea was overviewed and recommendations from the industry were introduced. It is strong opinion of Korea nuclear industry that nuclear policy should be changed from the support policy to the target management policy. In the point of reactor strategy, the life of light water reactor technology might be longer than expected before in Korea and it is emphasized that good maintenance of light water reactor technology and smooth transition program to the advanced technologies should be carefully considered. There are differences in the opinions between preferences to the evolutionary and/or passive, inherently safe reactors but, in the long-term point of view, it is judged to be desirable to have alternatives

Points of emphasis and objectives of reactor safety research

International Nuclear Information System (INIS)

Krewer, K.H.

1982-01-01

Reactor safety research is part of the presently running second programme on energy research and energy-engineering with which the Federal Government is connecting a whole bundle of economic and ecological aims: medium- and long-term assurance of energy supply, provision and efficient utilization of energy at favourable economic total costs, improvement of the technological performance, consideration of the requirements of the environmental protection, of the careful treatment of the resources, as well as of the protection of the population and personnel from the risks of conversion and use of energy. (orig.) [de

Oklo reactors and implications for nuclear science

OpenAIRE

Davis, E. D.; Gould, C. R.; Sharapov, E. I.

2014-01-01

We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlie...

Big Rock Point: 35 years of electrical generation

International Nuclear Information System (INIS)

Petrosky, T.D.

1998-01-01

On September 27, 1962, the 75 MWe boiling water reactor, designed and built by General Electric, of the Big Rock Point Nuclear Power Station went critical for the first time. The US Atomic Energy Commission (AEC) and the plant operator, Consumers Power, had designed the plant also as a research reactor. The first studies were devoted to fuel behavior, higher burnup, and materials research. The reactor was also used for medical technology: Co-60 radiation sources were produced for the treatment of more than 120,000 cancer patients. After the accident at the Three Mile Island-2 nuclear generating unit in 1979, Big Rock Point went through an extensive backfitting phase. Personnel from numerous other American nuclear power plants were trained at the simulator of Big Rock Point. The plant was decommissioned permanently on August 29, 1997 after more than 35 years of operation and a cumulated electric power production of 13,291 GWh. A period of five to seven years is estimated for decommissioning and demolition work up to the 'green field' stage. (orig.) [de

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)

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.

Evaluation of thermal physical properties for fast reactor fuels. Melting point and thermal conductivities

International Nuclear Information System (INIS)

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya

2006-10-01

Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (author)

Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

International Nuclear Information System (INIS)

Aydogdu, K.

1998-01-01

Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

Space reactor safety, 1985--1995 lessons learned

International Nuclear Information System (INIS)

Marshall, A.C.

1995-01-01

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration

Space reactor safety, 1985--1995 lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marshall, A.C.

1995-12-31

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

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.)

Gasification with nuclear reactor heat

International Nuclear Information System (INIS)

Weisbrodt, I.A.

1977-01-01

The energy-political ultimate aims for the introduction of nuclear coal gasification and the present state of technology concerning the HTR reactor, concerning gasification and heat exchanging components are outlined. Presented on the plans a) for hydro-gasification of lignite and for steam gasification of pit coal for the production of synthetic natural gas, and b) for the introduction of a nuclear heat system. The safety and environmental problems to be expected are portrayed. The main points of development, the planned prototype plant and the schedule of the project Pototype plant Nuclear Process heat (PNP) are specified. In a market and economic viability study of nuclear coal gasification, the application potential of SNG, the possible construction programme for the FRG, as well as costs and rentability of SNG production are estimated. (GG) [de

TU electric reactor model verification

International Nuclear Information System (INIS)

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

1989-01-01

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

Thermo-hydraulic simulations of the experimental fast reactor core

International Nuclear Information System (INIS)

Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

1985-01-01

A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sekine, Katsuhisa.

1987-01-01

Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

Development of integrated nuclear data utilization system for innovative reactors

International Nuclear Information System (INIS)

Naoki, Yamano; Masayuki, Igashira; Akira, Hasegawa; Kiyoshi, Kato

2005-01-01

An integrated nuclear data utilization system has been developing for innovative nuclear energy systems such as innovative reactors and accelerator-driven systems. The system has been constructed as a modular code system, which consists of a managing system and two subsystems. The management system named CONDUCT controls system resource management of the PC Linux server and the user authentication through Internet access. A subsystem is the nuclear data search and plotting subsystem based on a SPES engine developed by Hokkaido University. Nuclear data such as EXFOR, JENDL-3.3, ENDF/B-VI and JEFF-3.1 can be searched and plotted in the subsystem. The other is the nuclear data processing and utilization subsystem, which is able to handle JENDL-3.3, ENDF/B-VI and JEFF-3.1 to generate point-wise and group cross sections in several formats, and perform various criticality and shielding benchmarks for verification of nuclear data and validation of design methods for innovative reactors. This paper presents an overview of the integrated nuclear data utilization system, describes the progress of the system development to examine the operability of the user interface and discuss specifications of the two subsystems. (authors)

Dosimetry system of the RB reactor; Dozimetarski sistem reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Vukadin, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1962-07-01

Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor.

Boiling water reactor liquid radioactive waste processing system

International Nuclear Information System (INIS)

Anon.

1977-01-01

The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

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

Photocatalytic treatment of bioaerosols: impact of the reactor design.

Science.gov (United States)

Josset, Sébastien; Taranto, Jérôme; Keller, Nicolas; Keller, Valérie; Lett, Marie-Claire

2010-04-01

Comparing the UV-A photocatalytic treatment of bioaerosols contaminated with different airborne microorganisms such as L. pneumophila bacteria, T2 bacteriophage viruses and B. atrophaeus bacterial spores, pointed out a decontamination sensitivity following the bacteria > virus > bacterial spore ranking order, differing from that obtained for liquid-phase or surface UV-A photocatalytic disinfection. First-principles CFD investigation applied to a model annular photoreactor evidenced that larger the microorganism size, higher the hit probability with the photocatalytic surfaces. Applied to a commercial photocatalytic purifier case-study, the CFD calculations showed that the performances of the studied purifier could strongly benefit from rational reactor design engineering. The results obtained highlighted the required necessity to specifically investigate the removal of airborne microorganisms in terms of reactor design, and not to simply transpose the results obtained from studies performed toward chemical pollutants, especially for a successful commercial implementation of air decontamination photoreactors. This illustrated the importance of the aerodynamics in air decontamination, directly resulting from the microorganism morphology.

Measurement and Calculation of Gamma Radiation from HWZPR Reactor

International Nuclear Information System (INIS)

Jalali, Majid

2006-01-01

HWZPR is a research reactor with natural uranium fuel, D 2 O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1 H(n,γ) 2 H, 16 O(n,n'γ) 16 O, 2 H(n,γ) 3 H and 238 U(n,γ) 239 U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27 Al (n,γ) 28 Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3-11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious. (author)

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.

The spatial kinetic analysis of accelerator-driven subcritical reactor

International Nuclear Information System (INIS)

Takahashi, H.; An, Y.; Chen, X.

1998-02-01

The operation of the accelerator driven reactor with subcritical condition provides a more flexible choice of the reactor materials and of design parameters. A deep subcriticality is chosen sometime from the analysis of point kinetics. When a large reactor is operated in deep subcritical condition by using a localized spallation source, the power distribution has strong spatial dependence, and point kinetics does not provide proper analysis for reactor safety. In order to analyze the spatial and energy dependent kinetic behavior in the subcritical reactor, the authors developed a computation code which is composed of two parts, the first one is for creating the group cross section and the second part solves the multi-group kinetic diffusion equations. The reactor parameters such as the cross section of fission, scattering, and energy transfer among the several energy groups and regions are calculated by using a code modified from the Monte Carlo codes MCNPA and LAHET instead of the usual analytical method of ANISN, TWOTRAN codes. Thus the complicated geometry of the accelerator driven reactor core can be precisely taken into account. The authors analyzed the subcritical minor actinide transmutor studied by Japan Atomic Energy Research Institute (JAERI) using the code

RESEND, Infinitely Dilute Point Cross-Sections Calculation from ENDF/B Resonance Parameter. ADLER, ENDF/B Adler-Adler Resonance Parameter to Point Cross-Sections with Doppler Broadening

International Nuclear Information System (INIS)

Bhat, M.R.; Ozer, O.

1982-01-01

1 - Description of problem or function: RESEND generates infinitely- dilute, un-broadened, point cross sections in the ENDF format by combining ENDF File 3 background cross sections with points calculated from ENDF File 2 resonance parameter data. ADLER calculates total, capture, and fission cross sections from the corresponding Adler-Adler parameters in the ENDF/B File 2 Version II data and also Doppler-broadens cross sections. 2 - Method of solution: RESEND calculations are done in two steps by two separate sections of the program. The first section does the resonance calculation and stores the results on a scratch file. The second section combines the data from the scratch file with background cross sections and prints the results. ADLER uses the Adler-Adler formalism. 3 - Restrictions on the complexity of the problem: RESEND expects its input to be a standard mode BCD ENDF file (Version II/III). Since the output is also a standard mode BCD ENDF file, the program is limited by the six significant figure accuracy inherent in the ENDF formats. (If the cross section has been calculated at two points so close in energy that only their least significant figures differ, that interval is assumed to have converged, even if other convergence criteria may not be satisfied.) In the unresolved range the cross sections have been averaged over a Porter-Thomas distribution. In some regions the calculated resonance cross sections may be negative. In such cases the standard convergence criterion would cause an unnecessarily large number of points to be produced in the region where the cross section becomes zero. For this reason an additional input convergence criterion (AVERR) may be used. If the absolute value of the cross section at both ends of an interval is determined to be less than AVERR then the interval is assumed to have converged. There are no limitations on the total number of points generated. The present ENDF (Version II/III) formats restrict the total number of

Itese Newsletter, Number 23 - Autumn 2014

International Nuclear Information System (INIS)

Devezeaux de Lavergne, Jean-Guy; Gabriel, Sophie; Jasserand, Frederic; Le Duigou, Alain; Le Net, Elisabeth; Mansilla, Christine; Maziere, Dominique; Monnet, Antoine; Popiolek, Nathalie; Thais, Francoise; Yu, Julie

2014-01-01

The first article of this newsletter describes some aspects of the role of hydrogen in energy transition. It outlines the sensitivity of the production cost to the business model, the great number of applications of hydrogen as a chemical product and as an energy vector, the promising applications in mobility and transport, the good public image of this product, and the role hydrogen could play in energy transition. The second article comments the recent evolutions of the uranium market and outlines the perspective of concentration of mining companies, notably state companies. The third article proposes a comparative analysis of public policies in favour of solar photovoltaic development between Germany and China (in terms of objectives, of inputs and results, of production and installations). It also highlights the interactions between the photovoltaic policy strategies of both countries. The next article addresses the project development of two EPR nuclear reactors at Hinkley Point in the UK with the favourable recommendation of the European Competition Commission. Some brief news are then proposed (mainly about recent international meetings on energy and on uranium non conventional resources)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

Restoring Relationships

Science.gov (United States)

Umphrey, Jan

2013-01-01

Matthew R. Willis, an assistant principal at William C. Hinkley High School in Aurora, CO, is the 2013 NASSP/Virco National Assistant Principal of the Year. In this interview with "Principal Leadership," he is asked to relate what he believes are his largest accomplishments in his role as assistant principal at William C. Hinkley. He…

ADVANCED CONTROL FOR A ETHYLENE REACTOR

Directory of Open Access Journals (Sweden)

Dumitru POPESCU

2017-06-01

Full Text Available The main objective of this work is the design and implementation of control solutions for petrochemical processes, namely the control and optimization of a pyrolysis reactor, the key-installation in the petrochemical industry. We present the technological characteristics of this petrochemical process and some aspects about the proposed control system solution for the ethylene plant. Finally, an optimal operating point for the reactor is found, considering that the process has a nonlinear multi-variable structure. The results have been implemented on an assembly of pyrolysis reactors on a petrochemical platform from Romania.

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.

Effects of dew point on selective oxidation of TRIP steels containing Si, Mn, and B

Science.gov (United States)

Lee, Suk-Kyu; Kim, Jong-Sang; Choi, Jin-Won; Kang, Namhyun; Cho, Kyung-Mox

2011-04-01

The selective oxidation of Si, Mn, and B on TRIP steel surfaces is a widely known phenomenon that occurs during heat treatment. However, the relationship between oxide formation and the annealing factors is not completely understood. This study examines the effect of the annealing conditions (dew point and annealing temperature) on oxide formation. A low dew point of -40 °C leads to the formation of Si-based oxides on the surface. A high dew point of -20 °C changes the oxide type to Mn-based oxides because the formation of Si oxides on the surface is suppressed by internal oxidation. Mn-based oxides exhibit superior wettability due to aluminothermic reduction during galvanizing.

Generalized treatment of point reactor kinetics driven by random reactivity fluctuations via the Wiener-Hermite functional method

International Nuclear Information System (INIS)

Behringer, K.

1991-02-01

In a recent paper by Behringer et al. (1990), the Wiener-Hermite Functional (WHF) method has been applied to point reactor kinetics excited by Gaussian random reactivity noise under stationary conditions, in order to calculate the neutron steady-state value and the neutron power spectral density (PSD) in a second-order (WHF-2) approximation. For simplicity, delayed neutrons and any feedback effects have been disregarded. The present study is a straightforward continuation of the previous one, treating the problem more generally by including any number of delayed neutron groups. For the case of white reactivity noise, the accuracy of the approach is determined by comparison with the exact solution available from the Fokker-Planck method. In the numerical comparisons, the first-oder (WHF-1) approximation of the PSD is also considered. (author) 4 figs., 10 refs

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

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.

Thermosyphoning in the CANDU reactor

International Nuclear Information System (INIS)

Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

1984-01-01

Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

Present status of space nuclear reactor

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

The geology of some United Kingdom nuclear sites related to the disposal of low and medium level radioactive wastes

International Nuclear Information System (INIS)

Robins, N.S.

1980-06-01

The geological sequences beneath a further twelve nuclear sites in Britain are predicted from available data. Formations that are potentially suitable hosts for low and medium-level radioactive waste are identified and their relative merits assessed. Of the sites investigated, formations beneath six afford little or no potential, formations beneath a further 4 offer only moderate potential and sites underlain by the most favourable formations are Dungeness and Hinkley Point. (author)

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

Generating the flux map of Nigeria Research Reactor-1 for efficient ...

African Journals Online (AJOL)

One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

A novel knot selection method for the error-bounded B-spline curve fitting of sampling points in the measuring process

International Nuclear Information System (INIS)

Liang, Fusheng; Zhao, Ji; Ji, Shijun; Zhang, Bing; Fan, Cheng

2017-01-01

The B-spline curve has been widely used in the reconstruction of measurement data. The error-bounded sampling points reconstruction can be achieved by the knot addition method (KAM) based B-spline curve fitting. In KAM, the selection pattern of initial knot vector has been associated with the ultimate necessary number of knots. This paper provides a novel initial knots selection method to condense the knot vector required for the error-bounded B-spline curve fitting. The initial knots are determined by the distribution of features which include the chord length (arc length) and bending degree (curvature) contained in the discrete sampling points. Firstly, the sampling points are fitted into an approximate B-spline curve Gs with intensively uniform knot vector to substitute the description of the feature of the sampling points. The feature integral of Gs is built as a monotone increasing function in an analytic form. Then, the initial knots are selected according to the constant increment of the feature integral. After that, an iterative knot insertion (IKI) process starting from the initial knots is introduced to improve the fitting precision, and the ultimate knot vector for the error-bounded B-spline curve fitting is achieved. Lastly, two simulations and the measurement experiment are provided, and the results indicate that the proposed knot selection method can reduce the number of ultimate knots available. (paper)

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)

Neutronics issues in fusion-fission hybrid reactor design

International Nuclear Information System (INIS)

Liu Chengan

1995-01-01

The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

Point Lepreau nuclear project unlocks power future of New Brunswick

Energy Technology Data Exchange (ETDEWEB)

Burgess, E J [New Brunswick Electric Power Commission, Fredericton (Canada)

1976-08-01

Projects under development will increase the generating capacity of New Brunswick Electric Power Commission from 1.47 to 3.14 GW. Of these projects, the most significant is the 0.63 GW Point Lepreau CANDU reactor. Progress in the construction of the power reactor is summarized. The concrete reactor building was slipformed in April 1976. A second nuclear unit at Point Lepreau is being considered. Information is also provided on the new oil-fired station at Coleson Cove, a new oil- or coal-fired unit at Dalhousie and further hydro units at Mactaquac. A 0.16 GW pumped storage hydro unit at Green River is being considered. Information on transmission (Including the HVDC system), substations, and connection with Hydro Quebec is included.

Cross-section requirements for reactor neutron flux measurements from the user's point of view

International Nuclear Information System (INIS)

Mas, P.; Lloret, R.

1978-01-01

The dosimetry of testing materials irradiations involves in practice a lot of problems: fluences measurements, knowledge of spectrum, choice of a convenient set of cross section, damage rate determination, transposition from testing reactor to power reactor. From those problems, we consider that a temporary recommandation concerning the differential cross section of some fluence detectors is to be done, and that it is necessary to dispose of more accessible benchmarks in order to correlate cross section and computer codes. (author)

RHTF 2, a 1200 MWe high temperature reactor

International Nuclear Information System (INIS)

Brisbois, Jacques

1978-01-01

After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

Proposal of world network on material testing reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

2011-01-01

Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

CLINSULF sub-dew-point process for sulphur recovery

Energy Technology Data Exchange (ETDEWEB)

Heisel, M.; Marold, F.

1988-01-01

In a 2-reactor system, the CLINSULF process allows very high sulphur recovery rates. When operated at 100/sup 0/C at the outlet, i.e. below the sulphur solidification point, a sulphur recovery rate of more than 99.2% was achieved in a 2-reactor series. Assuming a 70% sulphur recovery in an upstream Claus furnace plus sulphur condenser, an overall sulphur recovery of more than 99.8% results for the 2-reactor system. This is approximately 2% higher than in conventional Claus plus SDP units, which mostly consist of 4 reactors or more. This means the the CLINSULF SSP process promises to be an improvement both in respect of efficiency and low investment cost.

The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

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.

SPV Analysis of CEDMCS in Advanced Power Reactors

Energy Technology Data Exchange (ETDEWEB)

Awwal, Arigi M.; Emmanuel, Efenji A. Emmanuel; Faragalla, Mohamed M.; Lee, Yong-kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

Single Point Vulnerability (SPV) is a component whose failure would directly cause an automatic or manual reactor scram or turbine trip. Although some power plants do not consider the cause of any reduction in power as SPV, others consider components that cause a reduction in power of as low as 2% as SPV. The Control Element Drive Mechanism Control System (CEDMCS) controls and regulates power supplied to drive the control rods with the Control Element Drive Mechanism (CEDM). A 4-coil CEDM is used in the newly built Advanced Power Reactor (APR) 1400 plant, while a new CEDMCS for 3-coil CEDM has been designed to be deployed to another APR1400 plant. This paper shows an approach to evaluate the SPVs that may be available in either of these two systems. System A design has employed a fail-safe concept to its design with less redundancies while System B design provides redundancy and design change although this comes at a high price for the Utility. The System B design has improved reliability but not necessarily eliminating the SPV items. Naturally, the cost of a new redundant system will be more. However, future work will examine the economic effect of the new system considering the operating experiences of power plants on the CEDMCS (i.e. SCRAM rates and power outage cost)

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.)

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

Energy Technology Data Exchange (ETDEWEB)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

2012-11-15

Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

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

Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

International Nuclear Information System (INIS)

Exarhos, C.A.; Van Swam, L.F.; Wahlquist, F.P.

1981-12-01

This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

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

Evolution of CANDU reactor design

International Nuclear Information System (INIS)

Pon, G.A.

1978-08-01

The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

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

Research reactor job analysis - A project description

International Nuclear Information System (INIS)

Yoder, John; Bessler, Nancy J.

1988-01-01

Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

Comparability of Results between Point-of-Care and Automated Instruments to Measure B-type Natriuretic Peptide

OpenAIRE

Shah, Kevin; Terracciano, Garrett J.; Jiang, Kevin; Maisel, Alan S.; Fitzgerald, Robert L.

2010-01-01

Objectives: Heart failure is one of the leading causes of death in the U.S. The incorporation of B-type natriuretic peptide (BNP) measurements when triaging patients presenting with shortness of breath has improved the diagnostic and prognostic ability of physicians. Currently, there are no point-of-care systems for quantifying BNP that can be used without sacrificing accuracy. We compared the analytical performance of the Abbott i-STAT analyzer, a handheld point-of-care system for measuring ...

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)

Power from plutonium: fast reactor fuel

International Nuclear Information System (INIS)

Bishop, J.F.W.

1981-01-01

Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

Empiric model for mean generation time adjustment factor for classic point kinetics equations

Energy Technology Data Exchange (ETDEWEB)

Goes, David A.B.V. de; Martinez, Aquilino S.; Goncalves, Alessandro da C., E-mail: david.goes@poli.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: alessandro@con.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

2017-11-01

Point reactor kinetics equations are the easiest way to observe the neutron production time behavior in a nuclear reactor. These equations are derived from the neutron transport equation using an approximation called Fick's law leading to a set of first order differential equations. The main objective of this study is to review classic point kinetics equation in order to approximate its results to the case when it is considered the time variation of the neutron currents. The computational modeling used for the calculations is based on the finite difference method. The results obtained with this model are compared with the reference model and then it is determined an empirical adjustment factor that modifies the point reactor kinetics equation to the real scenario. (author)

Empiric model for mean generation time adjustment factor for classic point kinetics equations

International Nuclear Information System (INIS)

Goes, David A.B.V. de; Martinez, Aquilino S.; Goncalves, Alessandro da C.

2017-01-01

Point reactor kinetics equations are the easiest way to observe the neutron production time behavior in a nuclear reactor. These equations are derived from the neutron transport equation using an approximation called Fick's law leading to a set of first order differential equations. The main objective of this study is to review classic point kinetics equation in order to approximate its results to the case when it is considered the time variation of the neutron currents. The computational modeling used for the calculations is based on the finite difference method. The results obtained with this model are compared with the reference model and then it is determined an empirical adjustment factor that modifies the point reactor kinetics equation to the real scenario. (author)

Thermohydraulic analysis of pressurized water reactors

International Nuclear Information System (INIS)

Veloso, M.A.

1980-01-01

The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

Power reactor noise studies and applications

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, V

2002-03-01

The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

Power reactor noise studies and applications

International Nuclear Information System (INIS)

Arzhanov, V.

2002-03-01

The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

Modular reactor head shielding system

International Nuclear Information System (INIS)

Jacobson, E. B.

1985-01-01

An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

Safety reviews of the Brazilian multipurpose reactor

International Nuclear Information System (INIS)

Soares, Humberto Vitor

2014-01-01

This work presents a model developed for thermal hydraulic (TH) simulation of the Multipurpose Brazilian Reactor (RMB), whose Brazilian proposal for design, construction and operation was established in 2007. This reactor has as main proposed the production of radioisotopes for use in exams of nuclear medicine, material tests and utilization of neutrons beams. Besides of the TH modeling and safety analysis of the reactor, the application of a methodology to perform coupled calculation thermal-hydraulic/neutron kinetic (TH/NK) is also presented. Initially, the RMB was modeled in the safety analysis RELAP5 code. This code performs the thermal hydraulic calculation using point kinetics. Subsequently, the model was adapted and verified to the RELAP5-3D© code. This code performs the process of internal coupling through the option of nodal neutron kinetics calculation using the NESTLE code which solves the neutron diffusion equation. To generate the neutronic group constants, which are macroscopic cross sections that serve as input data for the neutronic codes, it was used the WIMSD-5B cell calculation code. The neutron analysis code PARCS was also used to model the 3D RMB core in order to compare the results of radial and axial average power distribution with the results generated by RELAP5-3D© code and with the available results of the CITATION neutron kinetic code. The safety analyses demonstrated safe behavior of the reactor through situations of possible transients. The 3D coupled calculations to the steady state operation also showed expected behavior, as well as the RMB neutronic analyzes performed with the codes NESTLE and PARCS.(author)

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

Microdosimetric measurements in the thermal neutron irradiation facility of LENA reactor

International Nuclear Information System (INIS)

Colautti, P.; Moro, D.; Chiriotti, S.; Conte, V.; Evangelista, L.; Altieri, S.; Bortolussi, S.; Protti, N.; Postuma, I.

2014-01-01

A twin TEPC with electric-field guard tubes has been constructed to be used to characterize the BNCT field of the irradiation facility of LENA reactor. One of the two mini TEPC was doped with 50 ppm of 10 B in order to simulate the BNC events occurring in BNCT. By properly processing the two microdosimetric spectra, the gamma, neutron and BNC spectral components can be derived with good precision (∼6%). However, direct measurements of 10 B in some doped plastic samples, which were used for constructing the cathode walls, point out the scarce accuracy of the nominal 10 B concentration value. The influence of the Boral ® door, which closes the irradiation channel, has been measured. The gamma dose increases significantly (+51%) when the Boral ® door is closed. The crypt-cell-regeneration weighting function has been used to measure the quality, namely the RBE µ value, of the radiation field in different conditions. The measured RBE µ values are only partially consistent with the RBE values of other BNCT facilities. - Highlights: • A counter with two mini TEPCs, both equipped with electrical-field guard tubes, has been constructed. • The microdosimetric spectrum of the LENA-reactor irradiation vane has been studied. • The radiation-field quality (RBE) assessment confirms that the D n /D tot ratio is not an accurate parameter to characterize the BNCT radiation field