Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

International Nuclear Information System (INIS)

1978-09-01

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

1999-08-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

International Nuclear Information System (INIS)

Hastowo, Hudi; Tarigan, Alim

1999-01-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

The reactor antineutrino anomaly and low energy threshold neutrino experiments

Science.gov (United States)

Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

2018-01-01

Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

International Reactor Physics Experiment Evaluation (IRPhE) Project

International Nuclear Information System (INIS)

2013-01-01

The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

Experiences in stability testing of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Otaduy, P.J.

1986-01-01

The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Experiment operations plan for the MT-4 experiment in the NRU reactor

International Nuclear Information System (INIS)

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (II). The second experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-01-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-03-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Tritium experience in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

1998-01-01

Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

Energy Technology Data Exchange (ETDEWEB)

Gutmann, H

1972-06-05

The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

International Experience with Fast Reactor Operation & Testing

International Nuclear Information System (INIS)

Sackett, John I.; Grandy, C.

2013-01-01

Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

International Nuclear Information System (INIS)

Beech, D.J.; May, R.

2000-01-01

In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

Advanced test reactor testing experience-past, present and future

International Nuclear Information System (INIS)

Marshall, Frances M.

2006-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Gabaraev, B.A.

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

Detecting Dark Photons with Reactor Neutrino Experiments

Science.gov (United States)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ reactors as potential sources of intense fluxes of low-mass dark photons.

Storage experience in Hungary with fuel from research reactors

International Nuclear Information System (INIS)

Gado, J.; Hargitai, T.

1996-01-01

In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

Investigating the spectral anomaly with different reactor antineutrino experiments

Directory of Open Access Journals (Sweden)

C. Buck

2017-02-01

Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

Operational experience of the Marcoule reactors

International Nuclear Information System (INIS)

Conte, F.

1963-01-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

Leak before break experience in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Moan, G.D.; Coleman, C.E.

1988-04-01

The paper describes how the requirements for Leak-Before-Break are met in CANDU reactors. The requirements are based on operational and laboratory experience. After the onset of leakage in a fuel channel from a delayed hydride crack, time is available to the operator to take action before the crack grows to an unstable length. The time available is calculated using different models which use crack growth data from small specimen tests. When the results from crack growth behaviour experiments, carried out on components removed from reactor are used in the model, the time available for operator response is about 100 hours

Fast breeder reactors: can we learn from experience

International Nuclear Information System (INIS)

Keck, O.

1981-01-01

An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

Licensing experience of the HTR-10 test reactor

International Nuclear Information System (INIS)

Sun, Y.; Xu, Y.

1996-01-01

A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

Review of experiments for research reactors - approved 1974

International Nuclear Information System (INIS)

Anon.

1977-01-01

This standard establishes guidelines for the review and approval of experiments performed at research reactor facilities. This standard identifies the major areas that shall be reviewed for each experiment to ensure that it (a) falls within the limits delineated in the technical specifications, (b) does not present an unreviewed safety question as defined in 10 CFR Section 50.59 π2-, (c) does not constitute a threat to the health and safety of any individual or group of individuals, and (d) does not constitute a hazard to the reactor facility or other equipment. In addition, this standard recommends a system for classifying experiments to establish levels of review and approval commensurate with the level of risk inherent in the experiment

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

International Nuclear Information System (INIS)

Bess, John D.; Gullifor, Jim

2015-03-01

The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

Experience and prospects for developing research reactors of different types

International Nuclear Information System (INIS)

Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

2015-01-01

NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-01

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-15

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

Experience in operation of heavy water reactors

International Nuclear Information System (INIS)

Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

1999-01-01

The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

Experience in utilizing research reactors in Yugoslavia

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

1972-07-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Experience in utilizing research reactors in Yugoslavia

International Nuclear Information System (INIS)

Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

1972-01-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Development of the design of the High Temperature Gas Cooled Reactor experiment

International Nuclear Information System (INIS)

Lockett, G.E.; Huddle, R.A.U.

1960-01-01

Early in 1956 a small team was formed at the Atomic Energy Research Establishment, Harwell, to investigate the possibilities of the High Temperature Gas Cooled (H.T.G.C.) Reactor System. Although the primary objective of this team was to carry out a feasibility study of the system as a whole, it soon became apparent that, in addition to design studies and economic surveys of power producing reactors, the most appropriate approach to such a novel system was to carry out a design study of a relatively small (10 to 20 M.W.) Reactor Experiment, together with the necessary research and development work associated with such a reactor. This work proceeded within the U.K.A.E.A. during the three following years, and it was felt that realistic design proposals could be put forward with sufficient confidence to justify the detailed design and construction of a 20 M.W. Reactor Experiment. In April 1959 responsibility for this Reactor Experiment was taken over by the O.E.E.C. High Temperature Gas Cooled Reactor Project, the DRAGON Project, at the Atomic Energy Establishment, Winfrith, Dorset. In this Paper the research, development, and design work is reviewed, and the proposals for the Reactor Experiment are summarised. (author)

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

Grover, S. Blaine

2009-01-01

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

Non-standard interaction effects at reactor neutrino experiments

International Nuclear Information System (INIS)

Ohlsson, Tommy; Zhang, He

2009-01-01

We study non-standard interactions (NSIs) at reactor neutrino experiments, and in particular, the mimicking effects on θ 13 . We present generic formulas for oscillation probabilities including NSIs from sources and detectors. Instructive mappings between the fundamental leptonic mixing parameters and the effective leptonic mixing parameters are established. In addition, NSI corrections to the mixing angles θ 13 and θ 12 are discussed in detailed. Finally, we show that, even for a vanishing θ 13 , an oscillation phenomenon may still be observed in future short baseline reactor neutrino experiments, such as Double Chooz and Daya Bay, due to the existences of NSIs

Brookhaven Reactor Experiment Control Facility, a distributed function computer network

International Nuclear Information System (INIS)

Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

1975-11-01

A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

New safety experiments in decommissioned superheated steam reactor at Karlstein

International Nuclear Information System (INIS)

Koerting, K.

1986-01-01

This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

International Nuclear Information System (INIS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T.J.; Cesar, J.; Cushman, P.; Dent, J.B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H.R.; Hays, C.C.; Iyer, V.

2017-01-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

Energy Technology Data Exchange (ETDEWEB)

Agnolet, G.; Baker, W. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Barker, D. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Beck, R. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Carroll, T.J.; Cesar, J. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Cushman, P. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Dent, J.B. [Department of Physics, University of Louisiana at Lafayette, Lafayette, LA 70504 (United States); De Rijck, S. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Dutta, B. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Flanagan, W. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Fritts, M. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Gao, Y. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Department of Physics & Astronomy, Wayne State University, Detroit 48201 (United States); Harris, H.R.; Hays, C.C. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Iyer, V. [School of Physical Sciences, National Institute of Science Education and Research, Jatni - 752050 (India); and others

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

Change of neutron flow sensors effectiveness in the course of reactor experiments

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Kotov, V.M.; Zhotabaev, Zh.R.

2007-01-01

Full text: IGR reactor is a reactor of thermal capacity type. During the operation, uranium-graphite core can be heated up to 1500 deg. C and reactivity can be changed considerably. Core dimensions are comparatively small. Amount of control rods, providing required reactivity, is not big as well. Increasing of core temperature leads to the rise of neutrons path length in its basic material - graphite. Change of temperature is not even. All this causes the non-conservation of neutron flows ratio in irradiated sample and in the place of reactor power sensors installation. Deviations in this ratio were registered during the number of reactor experiments. Empiric corrections can be introduced in order to decrease influence of change of neutron flow effectiveness upon provision of required parameters of investigated matters load. However, dependence of these corrections upon many factors can lead to the increasing of instability of process control. Previous experiment-calculated experiments showed inequality of neutron field in the place of sensors location (up to tens of percent), low effectiveness of experimental works, carried out without access to the individual reactor laying elements. Imperfection during the experiment was an idea of possibility to connect distribution of out of reactor neutron flow and control rods position. Subsequent analysis showed that for the development of representative phenomenon model it is necessary to take into account reactor operation dynamic subject to unevenness of heating of individual laying parts. Elemental calculations showed that temperature laying effects in the change of neutron outer field are great. Algorithm of calculations for the change of outer filed and field of investigated fabrication includes calculation of neutron-physic reactor characteristics interlacing with calculations of thermal-physic reactor characteristics, providing correlation of temperature fields for neutron-physic calculations. In the course of such

DOE's foreign research reactor transportation services contract: Perspective and experience

International Nuclear Information System (INIS)

Patterson, John

1997-01-01

DOE committed to low- and moderate-income countries participating in the foreign research reactor spent fuel returns program that the United States government would provide for the transportation of the spent fuel. In fulfillment of that commitment, DOE entered into transportation services contracts with qualified, private-sector firms. NAC will discuss its experience as a transportation services provider, including range of services available to the foreign reactors, advantages to DOE and to the foreign research reactors, access to contract services by high income countries and potential advantages, and experience with initial tasks performed under the contract. (author)

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

Energy Technology Data Exchange (ETDEWEB)

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance, (France); Snoj, Luka [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Tarchalski, Mikolaj [National Centre for Nuclear Research, ulica Andrzeja Soltana 7, 05-400 Otwock (Swierk), (Poland); Dewynter-Marty, Veronique [CEA, DEN, DANS, DRSN, SIREN, LESCI, Saclay, F-91191 Gif sur Yvette, (France); Malouch, Fadhel [CEA, DEN, DANS, DM2S, SERMA, Saclay, F-91191 Gif sur Yvette, (France)

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

Present and future oscillation experiments at reactors

International Nuclear Information System (INIS)

Mikaehlyan, L.A.

2001-01-01

A report is presented on recent progress and developments (since the NANP'99 Conference) in the current and future long baseline (∼100 - 800 km) oscillation experiments at reactors. These experiments, under certain assumptions, can fully reconstruct the internal mass structure of the electron neutrino and provide a laboratory test of solar and atmospheric neutrino problems

Liquid jet experiments: relevance to inertial confinement fusion reactors

International Nuclear Information System (INIS)

Hoffman, M.A.

1981-01-01

In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

Physics experiment on the Dragon reactor

Energy Technology Data Exchange (ETDEWEB)

Hunt, C.

1974-10-15

The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

LMFBR plant parameters

International Nuclear Information System (INIS)

1985-07-01

This document has been prepared on the basis of information compiled by the members of the IAEA International Working Group on Fast Reactors (IWGFR). It contains parameters of 25 experimental, prototype and commercial size liquid metal fast breeder reactors (LMFBR). Most of the reactors are currently in operation, under construction or in an advanced planning stage. Parameters of the Clinch River Breeder Reactor (USA) are presented because its design was nearly finished and most of the components were fabricated at the time when the project was terminated. Three reactors (RAPSODIE (France), DFR (UK) and EFFBR (USA)) have been shut down. However, they are included in the report because of their important role in the development of LMFBR technology from first LMFBRs to the prototype size fast reactors. The first LMFBRs (CLEMENTINE (USA), EBR-1 (USA), BR-2 (USSR), BR-5 (USSR)) and very special reactors (LAMPRE (USA), SEFOR (USA)) were not recommended by the members of the IWGFR to be included in the report

Dry cooling tower operating experience in the LOFT reactor

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

Lessons from feedback of safety operating experience for reactor physics

International Nuclear Information System (INIS)

Suchomel, J.; Rapavy, S.

1999-01-01

Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

Experience from and research activities at the Otaniemi TRIGA reactor

International Nuclear Information System (INIS)

Bars, Bruno

1976-01-01

Experience from the Finnish TRIGA Reactor is reported, small changes and improvements in the control console of the Fir-1 reactor have been made. A minicomputer based data collecting system is planned and installed. It will be used for collecting data from operation and radiation monitors including the new isotope laboratory, and also simultaneously smaller experiments such as control rod calibration. A minicomputer is used for on-line reactor noise studies. The automatic uranium analyzer has a maximum sensitivity of 0.03 μg U 235 and 1.2 Th 232 . The system is now used at a sampling rate of about one sample per minute. (author)

Power cycling experiments in INR-TRIGA-SSR Reactor

International Nuclear Information System (INIS)

Dumitru, M.

2008-01-01

The in-reactor experimental program started this summer with some power cycling experiments to provide date on fuel behaviour under abnormal reactor operating conditions. The paper describes the irradiation device, its operational features and an original 'under-flux' movement system. Also, there are presented main data of irradiation device (pressure, flow, temperature, construction), in-pile section, location, sample, instrumentation, experimental sequences and operating data of Interest for the experimenters. (author)

Experiments with the SUR 100 training reactor

International Nuclear Information System (INIS)

Milicic, B.

1984-06-01

This paper contains a compilation of various experiments using the SUR - 100 reactor for training purposes, which have been widly proved in practical work at the School for Nuclear Technology of the Karlsruhe Research Center. (orig.) [de

Reactor physics experiment plan using TCA

Energy Technology Data Exchange (ETDEWEB)

Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors, which aims at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. This report is to plan critical experiments using TCA in JAERI. Critical Experiments performed so far in Europe and Japan are reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX-fuel rods used in the experiments is obtained by calculations and modification of the equipment for the experiments are shown. New MOX fuel and UO{sub 2} fuel rods are necessary for the RMWR critical experiments. Number of MOX fuel rods is 1000 for Plutonium fissile enrichment of 5 wt%, 1000 for 10 wt%, 1500 for 15 wt% and 500 for 20 wt%, respectively. Depleted UO{sub 2} fuel rods for blanket/buffer region are 4000. Driver fuel rods of 4.9 wt% UO{sub 2} are 3000. Modification of TCA facility is requested to treat the large amount of MOX fuels from safety point of view. Additional shielding device at the top of the tank for loading the MOX fuels and additional safety plates to ensure safety are requested. The core is divided into two regions by inserting an inner tank to avoid criticality in MOX region only. The test region is composed by MOX fuel rods in the inner tank. Criticality is established by UO{sub 2} driver fuel rods outside of the inner tank. (Tsuchihashi, K.)

Predictions on an HTR coolant composition after operational experience with experimental reactors

International Nuclear Information System (INIS)

Nieder, R.

1981-01-01

Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

International Nuclear Information System (INIS)

Moiseyev, A.V.

2008-01-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

2008-07-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

International Nuclear Information System (INIS)

Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

Physics experiments with the operating reactor

Energy Technology Data Exchange (ETDEWEB)

Cullington, G R; King, D C

1973-09-27

Experimental techniques have been developed and used on Dragon to give consistent information on excess reactivity and shut down margin. The reactivity measurements have been correlated with the theoretical calculations and have led to improvements in the calculations. The methods used and the results obtained are accepted by the Safety Committee as sufficient evidence for compliance with the fuel loading safety rules. Although the reactor was not designed as an experimental facility, flux and dose measurements experiments have been successfully carried out. Mass flow and negative reactivity transient measurements have been carried out. These are valuable for demonstration of the flexibility of the reactor system and for giving confidence in theoretical calculations.

The double chooz reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Botella, I Gil [CIEMAT, Basic Research Department, Avenida Complutense, 22, 28040 Madrid (Spain)], E-mail: ines.gil@ciemat.es

2008-05-15

The Double Chooz reactor neutrino experiment will be the next detector to search for a non vanishing {theta}{sub 13} mixing angle with unprecedented sensitivity, which might open the way to unveiling CP violation in the leptonic sector. The measurement of this angle will be based in a precise comparison of the antineutrino spectrum at two identical detectors located at different distances from the Chooz nuclear reactor cores in France. Double Chooz is particularly attractive because of its capability to measure sin{sup 2} (2{theta}{sub 13}) to 3{sigma} if sin{sup 2}(2{theta}{sub 13}) > 0.05 or to exclude sin{sup 2}(2{theta}{sub 13}) down to 0.03 at 90% C.L. for {delta}m{sup 2} = 2.5 x 10{sup -3} eV{sup 2} in three years of data taking with both detectors. The construction of the far detector starts in 2008 and the first neutrino results are expected in 2009. The current status of the experiment, its physics potential and design and expected performance of the detector are reviewed.

Substantiation of physical concepts of fast reactors in Russia: experience and prospects

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

2008-07-01

The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

U-233 fuelled low critical mass solution reactor experiment PURNIMA II

International Nuclear Information System (INIS)

Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

1987-01-01

A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

International Nuclear Information System (INIS)

Joo, Kyung Kwang; Shin, Chang Dong

2014-01-01

For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ 13 , providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ 12 ,Δm 12 2 , and mass hierarchy will be explored. The precise measurement of θ 13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Liquid metal cooled reactors: Experience in design and operation

International Nuclear Information System (INIS)

2007-12-01

on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

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

The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

International Nuclear Information System (INIS)

Junkrans, S.; Vesterlund, G.; Vaernild, O.

1980-01-01

The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

International Nuclear Information System (INIS)

Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

2011-01-01

The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

International Nuclear Information System (INIS)

Higgins, J.; Nasta, K.

1997-01-01

The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs

Experience with valves for PHWR reactors

International Nuclear Information System (INIS)

Narayan, K.; Mhetre, S.G.

1977-01-01

Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

US graphite reactor D ampersand D experience

International Nuclear Information System (INIS)

Garrett, S.M.K.; Williams, N.C.

1997-02-01

This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE)

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Baek, W. P.; Chung, M. K.

2007-06-01

The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

Nuclear data and integral experiments in reactor physics

International Nuclear Information System (INIS)

Farinelli, U.

1980-01-01

The material given here broadly covers the content of the 10 lectures delivered at the Winter Course on Reactor Theory and Power Reactors, ICTP, Trieste (13 February - 10 March 1978). However, the parts that could easily be found in the current literature have been omitted and replaced with the appropriate references. The needs for reactor physics calculations, particularly as applicable to commercial reactors, are reviewed in the introduction. The relative merits and shortcomings of fundamental and semi-empirical methods are discussed. The relative importance of different nuclear data, the ways in which they can be measured or calculated, and the sources of information on measured and evaluated data are briefly reviewed. The various approaches to the condensation of nuclear data to multigroup cross sections are described. After some consideration to the sensitivity calculations and the evaluation of errors, some of the most important type of integral experiments in reactor physics are introduced, with a view to showing the main difficulties in the interpretation and utilization of their results and the most recent trends in experimentation. The conclusions try to assign some priorities in the implementation of experimental and calculational capabilities, especially for a developing country. (author)

Decommissioning experience of the Japan power demonstration reactor

International Nuclear Information System (INIS)

Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

1992-01-01

Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

Chasing {theta}{sub 13} with new reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Lasserre, Th. [DSM/DAPNIA/SPP, CEA/Saclay, 91191 Gif-sur-Yvette (France)

2005-12-15

It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the {theta}{sub 13} mixing angle, free from any parameter degeneracies and correlations induced by matter effect and the unknown leptonic Dirac CP phase. The current best constraint on the third mixing angle comes from the Chooz reactor neutrino experiment sin{sup 2}(2{theta}{sub 13})<0.2 (90 % C.L., {delta}m{sub atm}{sup 2}=2.010{sup -3} eV{sup 2}). Several projects of experiment, with different timescales, have been proposed over the last two years all around the world. Their sensitivities range from sin{sup 2}(2{theta}{sub 13})<0.01 to 0.03, having thus an excellent discovery potential of the {nu}{sub e} fraction of {nu}{sub 3}.

Daya bay reactor neutrino experiment

International Nuclear Information System (INIS)

Cao Jun

2010-01-01

Daya Bay Reactor Neutrino Experiment is a large international collaboration experiment under construction. The experiment aims to precisely determine the neutrino mixing angle θ 13 by detecting the neutrinos produced by the Daya Bay Nuclear Power Plant. θ 13 is one of two unknown fundamental parameters in neutrino mixing. Its magnitude is a roadmap of the future neutrino physics, and very likely related to the puzzle of missing antimatter in our universe. The precise measurement has very important physics significance. The detectors of Daya Bay is under construction now. The full operation is expected in 2011. Three years' data taking will reach the designed the precision, to determine sin 2 2θ 13 to better than 0.01. Daya Bay neutrino detector is an underground large nuclear detector of low background, low energy, and high precision. In this paper, the layout of the experiment, the design and fabrication progress of the detectors, and some highlighted nuclear detecting techniques developed in the detector R and D are introduced. (author)

Operating Experience from Events Reported to the IAEA Incident Reporting System for Research Reactors

International Nuclear Information System (INIS)

2015-03-01

Operating experience feedback is an effective mechanism in providing lessons learned from events and the associated corrective actions to prevent them, helping to improve safety at nuclear installations. The Incident Reporting System for Research Reactors (IRSRR), which is operated by the IAEA, is an important tool for international exchange of operating experience feedback for research reactors. The IRSRR reports contain information on events of safety significance with their root causes and lessons learned which help in reducing the occurrence of similar events at research reactors. To improve the effectiveness of the system, it is essential that national organizations demonstrate an appropriate interest for the timely reporting of events important to safety and share the information in the IRSRR database. At their biennial technical meetings, the IRSRR national coordinators recommended collecting the operating experience from the events reported to the IRSRR and disseminating it in an IAEA publication. This publication highlights the root causes, safety significance, lessons learned, corrective actions and the causal factors for the events reported to the IRSRR up to September 2014. The publication also contains relevant summary information on research reactor events from sources other than the IRSRR, operating experience feedback from the International Reporting System for Operating Experience considered relevant to research reactors, and a description of the elements of an operating experience programme as established by the IAEA safety standards. This publication will be of use to research reactor operating organizations, regulators and designers, and any other organizations or individuals involved in the safety of research reactors

The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

International Nuclear Information System (INIS)

Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

2010-01-01

LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

An Overview of the International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

Briggs, J. Blair; Gulliford, Jim

2014-01-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

Fast breeder reactors: Experience and trends. V. 2

International Nuclear Information System (INIS)

1986-01-01

The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

Operational experience of decommissioning techniques for research reactors in the United Kingdom

International Nuclear Information System (INIS)

England, M.R.; McCool, T.M.

2002-01-01

In previous co-ordinated research projects (CRP) conducted by the IAEA no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. As experience was gained and technology advanced it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. It was within this context that a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001. This paper considers the experience gained from the decommissioning of two research reactors during the course of the CRP namely: (a) the ICI Triga Mk I reactor at Billingham UK which was largely complete by the end of the research project and (b) the Argonaut 100 reactor at the Scottish Universities Research and Reactor centre at East Kilbride in Scotland which is currently is the early stages of dismantling/site operations. It is the intention of this paper with reference to the two case studies outlined above to compare the actual implementation of these works against the original proposals and identify areas that were found to be problematical and/or identify any lessons learnt. (author)

Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

The primary circuit of the dragon high temperature reactor experiment

International Nuclear Information System (INIS)

Simon, R.

2005-01-01

The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

Application of the Dragon reactor experiment to the safety evaluation of current HTR systems

International Nuclear Information System (INIS)

Ashworth, F.P.O.; Faircloth, R.L.

1976-01-01

An important component of the confidence required for the safety assessment of high-temperature reactors is the experimental proof of phenomena such as fission product release or core corrosion. The most convincing experiments are those carried out in a reactor. This paper outlines the scope of experiments relevant to safety which can be done in the Dragon Reactor Experiment and describes as an example the experimental campaign and the current outcome of the work on validating the predictions of caesium release and migration. (author)

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

FELIX experiments and computational needs for eddy current analysis of fusion reactors

International Nuclear Information System (INIS)

Turner, L.R.

1984-01-01

In a fusion reactor, changing magnetic fields are closely coupled to the electrically-conducting metal structure. This coupling is particularly pronounced in a tokamak reactor in which magnetic fields are used to confine, stabilize, drive, and heat the plasma. Electromagnetic effects in future fusion reactors will have far-reaching implications in the configuration, operation, and maintenance of the reactors. This paper describes the impact of eddy-current effects on future reactors, the requirements of computer codes for analyzing those effects, and the FELIX experiments which will provide needed data for code validation

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

High-burn-up fuels for fast reactors. Past experience and novel applications

International Nuclear Information System (INIS)

Weaver, Kevan D.; Gilleland, John; Whitmer, Charles; Zimmerman, George

2009-01-01

Fast reactors in the U.S. routinely achieved fuel burn-ups of 10%, with some fuel able to reach peak burn-ups of 20%, notably in the Experimental Breeder Reactor II and the Fast Flux Test Facility. Maximum burn-up has historically been constrained by chemical and mechanical interactions between the fuel and its cladding, and to some extent by radiation damage and thermal effects (e.g., radiation-induced creep, thermal creep, and radiation embrittlement) that cause the cladding to weaken. Although fast reactors have used several kinds of fuel - including oxide, metal alloy, carbide, and nitride - the vast majority of experience with fast reactors has been using oxide (including mixed oxide) and metal-alloy fuels based on uranium. Our understanding of high-burn-up operation is also limited by the fact that breeder reactor programs have historically assumed that their fuel would eventually undergo reprocessing; the programs thus have not made high burn-up a top priority. Recently a set of novel designs have emerged for fast reactors that require little initial enrichment and no reprocessing. These reactors exploit a concept known as a traveling wave (sometimes referred to as a breed-and-burn wave, fission wave, or nuclear-burning wave). By breeding and using its own fuel in place as it operates, a traveling-wave reactor can obtain burn-ups that approach 50%, well beyond the current base of knowledge and experience. Our computational work on the physics of traveling-wave reactors shows that they require metal-alloy fuel to provide the margins of reactivity necessary to sustain a breed-and-burn wave. This paper reviews operating experience with high-burn-up fuels and the technical feasibility of moving to a qualitatively new burn-up regime. We discuss our calculations on traveling-wave reactors, including those concerning the possible use of thorium. The challenges associated with high burn-up and fluence in fuels and materials are also discussed. (author)

Reactor dynamics experiment of N.S. Mutsu using pseudo random signal. 1

International Nuclear Information System (INIS)

Hayashi, Koji; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Shimazaki, Junya; Inoue, Kimihiko; Ochiai, Masaaki.

1993-10-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, reactor noise experiments using pseudo random binary sequences (PRBS) have been planned, and a preliminary experiment was performed on March 4, 1991 in the first experimental navigation with the aim of checking the experimental procedures and conditions. The experiments using both reactivity and load disturbances were performed at 70 % of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. From the results obtained, we confirmed that (1) the procedures and experimental conditions determined prior to the experiment were suitable for performing the PRBS experiments, (2) when the PRBS disturbances were applied, the plant state remained quite stable, and (3) the quality of the measured data is adequate for the purpose of dynamics analysis. This paper summarizes the planning and preparation of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Operating experience and maintenance at the TRIGA Mark II LENA reactor

International Nuclear Information System (INIS)

Cingoli, F.; Meloni, S.; Alloni, L.

1986-01-01

A summary of reactor operation and maintenance in the time period 1982-1986 is presented and discussed. Some problems occurred from instrumentated aluminum cladded elements. Both of them presented damage in the cable tubes and one element showed a protuberance in the cladding. They were replaced with stainless - steel cladded ones. Both elements were sealed up in stainless - steel tubes and put away in wells, 3 meters deep, in the reactor room floor. Some minor problems, correlated to the quite aid instrumentation of the console, are reported. The reactor activity in the last four years was conditioned by the developing of the n - n-bar oscillation NADIR experiment. The thermal column was dismantled and rebuilt in consideration of the Nadir experiment necessities and this job is described in detail. The building containing, the target and the void pipe, presented in 1982 Conference, are now completely operating and the experiment is running. (author)

A review of experiments and results from the transient reactor test (TREAT) facility

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

International Nuclear Information System (INIS)

1975-12-01

The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

Two-phase reduced gravity experiments for a space reactor design

International Nuclear Information System (INIS)

Antoniak, Z.I.

1986-08-01

Future space missions envision the use of large nuclear reactors utilizing either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed, to coordinate all ongoing and planned reduced gravity flow experiments

Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

International Nuclear Information System (INIS)

Penndorf, K.

1976-04-01

Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

New Reactor Siting, Licensing and Construction Experience. Proceedings of the 2. CNRA International Workshop on 'New Reactor Siting, Licensing and Construction Experience'

International Nuclear Information System (INIS)

2013-01-01

This report documents the proceedings from the 2. Workshop on New Reactor Siting, Licensing and Construction Experience. A total of 45 specialists from 16 countries and international organisations attended. The meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the US Nuclear Regulatory Commission (U.S.NRC). The objectives of the workshop were to provide a forum to exchange information on lessons learned from siting, licensing and constructing new nuclear power plants around the world. Key focus areas included siting practices and regulatory positions that have been enhanced as a result of the Fukushima accident; lessons learned from licensing and design review approaches and challenges, construction experience and recommendations for regulatory oversight; and regulatory cooperation on generic and design specific issues through the MDEP specific working groups. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to regulatory cooperation on generic and design specific issues, MDEP working groups (EPR, AP1000), vendor inspection co-operation, digital I and C, and codes and standards. The second technical session was intended to discuss and share regulatory positions on siting practices and enhancements as a result of lessons learned from Fukushima accident. The third technical session addressed the construction experience and regulatory oversight of new reactor construction activities. And the fourth technical session included presentations on the lessons learned from regulatory licensing reviews of new reactor designs

State system experience with safeguarding power reactors

International Nuclear Information System (INIS)

Roehnsch, W.

1982-01-01

This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

The Role of Non-Destructive Testing in the Los Alamos Reactor Programme; Role des Essais Non Destructifs dans le Programme de Reacteurs de los Alamos; Rol' nedestruktivnykh ispytanij materialov v Los-Alamosskoj reaktornoj programme; Papel de los Metodos de Ensayo No Destructivo en el Programa de Reactores de Los Alamos

Energy Technology Data Exchange (ETDEWEB)

Tenney, G. H. [University of California, Los Alamos Scientific Laboratory, Los Alamos, NM (United States)

1965-10-15

The Los Alamos scientific Laboratory, operated Dy me University of california for me united states Atomic Energy Commission, has been active for more than twenty years in developing, designing, and building nuclear reactors of four general types: research, power, rocket propulsion, and critical assembly. The Non-destructive Testing Group serves practically all the activities and projects of the Laboratory; this paper describes some of the unique non-destructive testing techniques and applications developed for and used in the reactor programme. LAPRE (Los Alamos Power Reactor Experiment) was based on the use of a uranium phosphate solution at high temperature. This solution is very corrosive, and all parts in contact with it were clad with gold. Special radiographic techniques were used to inspect the gold during the process of producing rolled sheet from cast ingot. The welded seams were similarly inspected. An electrode-potential inspection method was developed for checking the gold surfaces for imbedded impurities. The fundamental concept of LAMPRE (Los Alamos Molten Plutonium Reactor Experiment) is the use of liquid - rather than solid - plutonium metal as fuel. Tantalum capsules contained the fuel. Novel nondestructive testing methods were used to check the soundness of base metal and welds during the production of the capsules, and to study the plutonium-loaded capsules before, during, and after melt-freeze tests. A molten plutonium pump experiment was followed with radiographic techniques, including a gamma-ray closed television circuit. For UHTREX (Ultra High Temperature Reactor Experiment), now under construction, micro radiographic and electron microscopic studies have been made on 150-{mu}m-diam. pyrocarbon-coated uranium carbide beads, to evaluate uranium migration as a function of temperature. The amount, and uniformity, of the uranium loading in the UHTREX graphite elements are determined with specially designed scintillation counters. About 90% of

Operational experience of the Marcoule reactors; Experience d'exploitation des reacteurs de Marcoule

Energy Technology Data Exchange (ETDEWEB)

Conte, F [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1963-07-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [French] Les resultats atteints apres trois ans de fonctionnement des reacteurs G-2/G-3 permettent une accumulation considerable de l'experience d'exploitation de ces reacteurs. Les principales originalites: - caisson en beton precontraint - chargement en marche - surveillance automatique des temperatures sont largement justifiees par l'exploitation actuelle. L'auteur confirme l'interet de ces solutions d'avant-garde et en tire des conclusions pour les etudes de futures centrales nucleaires. (auteur)

Quality assurance program plan for the Reactor Research Experiment Programs (RREP)

International Nuclear Information System (INIS)

Pipher, D.G.

1982-05-01

This document describes the Quality Assurance Program plans which will be applied to tasks on Reactor Research Experiments performed on Sandia National Laboratories' reactors. The program provides for individual project or experiment quality plan development and allows for reasonable plan flexibility and maximum plan visibility. Various controls and requirements in this program plan are considered mandatory on all features which are identified as important to public health and safety (Level I). It is the intent of this document that the Quality Assurance program comprise those elements which will provide adequate assurance that all components, equipment, and systems of the experiments will perform as designed, and hence prevent delays and costs due to rejections or failures

Combined potential of future long-baseline and reactor experiments

International Nuclear Information System (INIS)

Huber, P.; Lindner, M.; Rolinec, M.; Schwetz, T.; Winter, W.

2005-01-01

We investigate the determination of neutrino oscillation parameters by experiments within the next ten years. The potential of conventional beam experiments (MINOS, ICARUS, OPERA), superbeam experiments (T2K, NOνA), and reactor experiments (D-CHOOZ) to improve the precision on the 'atmospheric' parameters Δm 31 2 , θ 23 , as well as the sensitivity to θ 13 are discussed. Further, we comment on the possibility to determine the leptonic CP-phase and the neutrino mass hierarchy if θ 13 turns out to be large

IRPhE - International Reactor Physics Experiments database

International Nuclear Information System (INIS)

Sartori, E.

2004-01-01

The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists and has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. It is a significant saving results from disseminating a standard benchmark set to be used worldwide. A framework for professionals that use the standard benchmark set to validate and verify modeling codes and data for radiation transport, criticality safety and reactor physics applications guarantees a comparative set of analyses. It represents also a good basis for pinpointing important gaps and where efforts should be concentrated and ensures knowledge and competence preservation, management and transfer in nuclear science and engineering. A large number of experimentalists, physicists, evaluators, modelers have devoted large amounts of their efforts and competencies to produce the data on which the methods we are using today are based. These data are far from having been exploited fully for the different nuclear and radiation technologies. This wealth of information needs to be preserved in a form more easily exploitable by modern information technology and for use in connection with novel and refined computational models with limitations of the past removed. These data will form the basis for the studies of more advanced nuclear technology, will be instrumental in identifying areas where there is a lack of knowledge and thus provide support to justifying new experiments that would reduce design uncertainties and consequently costs. Improvement of

Maximizing the use of research reactors in training power reactor operating staff with special reference to US experience

International Nuclear Information System (INIS)

Cox, J.A.

1976-01-01

Research reactors have been used in training nuclear power plant personnel for many years. Using the experience in the United States of America a programme is proposed that will maximize the training conducted at a research reactor and lessen the time that the staff must spend training elsewhere. The programme is adaptable to future training of replacement staff and for staff retraining. (author)

Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

International Nuclear Information System (INIS)

Salvatores, M.

1986-09-01

Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

System Requirements Document for the Molten Salt Reactor Experiment

Energy Technology Data Exchange (ETDEWEB)

Aigner, R.D.

2000-04-01

The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

French experience concerning expansion compensating devices on the primary systems of nuclear reactors

International Nuclear Information System (INIS)

Vrillon, B.; Raynal, A.

1980-01-01

French experience in the use of large expansion bellows in the presence of hot sodium is extremely limited. This stems from the 'pool' structure of the primary circuit, adopted in France to eliminate the need to solve expansion problems affecting the primary piping of loop reactors. Furthermore, until the present time, the use of bellows on secondary circuits has neither been implemented nor considered. A few bellows nevertheless exist on the Phenix and Super-Phenix reactors, and these perform separation functions, for example, between sodium at different temperature and/or pressures, or tightness functions in gaseous environment at the component penetrations in the slabs. The dimension criteria applied to these bellows are the general rules for structural dimensioning. Since they do not form part of a circuit wall, they do not need to be discussed. Note, however, that these components have not raised any particular problems thus far. Expansion bellows exist in France on the primary circuits of certain nuclear reactors of the natural uranium/graphite/gas type. These reactors have been in operation for many years, and some lessons can be drawn from this experience in the use of bellows in representative conditions on power reactor circuits. Liquid sodium raises specific problems with respect to circuit operation and material behavior. However, many problems in the use of bellows are independent of the fluid conveyed in the circuits. This is why the experience gained with gas type power reactors appears to be useful in considering the possible future use of bellows on sodium reactor circuits

Plant experience of experimental fast reactor 'Joyo'

International Nuclear Information System (INIS)

1982-01-01

The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

French gas cooled reactor experience with moisture ingress

International Nuclear Information System (INIS)

Bastien, D.; Brie, M.

1995-01-01

During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

Experience with mechanical segmentation of reactor internals

International Nuclear Information System (INIS)

Carlson, R.; Hedin, G.

2003-01-01

Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

Experiences with fast breeder reactor education in laboratory and short course settings

International Nuclear Information System (INIS)

Waltar, A.E.

1983-01-01

The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

The SM and MIR reactors operation experience

International Nuclear Information System (INIS)

Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

1995-01-01

The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

Further experience in simulation of rod drop experiments in the Loviisa and Mochovce reactors

International Nuclear Information System (INIS)

Siltanen, P.; Kaloinen, E.; Tanskanen, A.; Mattila, R.

2001-01-01

Simulations of reactor scram experiments using the 3-dimensional kinetics code HEXTRAN have been updated for the initial cores of Loviisa-1 and 2 Mochovce-1 and have been extended to burned cores of Loviisa-1. In these simulations, the entire experiment is simulated dynamically, including the behaviour of the core, the signal of the ionization chamber, and the inverse point kinetics of the reactivity meter. The predicted output of the reactivity meter is compared with the output observed during the experiment (Authors)

Test on the reactor with the portable digital reactivity meter for physical experiment

International Nuclear Information System (INIS)

Huang Liyuan

2010-01-01

Test must be performed on the zero power reactor During the development of portable digital reactivity meter for physical experiment, in order to check its measurement function and accuracy. It describes the test facility, test core, test methods, test items and test results. The test results show that the instrument satisfy the requirements of technical specification, and satisfy the reactivity measurement in the physical experiments on reactors. (authors)

Description of the french graphite reactor and of the experiments performed in 1956

International Nuclear Information System (INIS)

Bussac, J.; Leduc, C.; Zaleski, C.P.

1957-01-01

This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author) [fr

Latest Results from the Daya Bay Reactor Neutrino Experiment

CERN Multimedia

CERN. Geneva

2014-01-01

Among all the fundamental particles that have been experimentally observed, neutrinos remain one of the least understood. The Daya Bay Reactor Neutrino Experiment in China consists of eight identical detectors placed underground at different baselines from three groups of nuclear reactors, a configuration that is ideally suited for studying the properties of these elusive particles. This talk will present three sets of results that have just recently been released by the Daya Bay Collaboration: (i) a precision measurement of the oscillation parameters that drive the disappearance of electron antineutrinos at short baselines, (ii) a search for sterile neutrino mixing, and (iii) a high-statistics determination of the absolute flux and spectrum of reactor-produced electron antineutrinos. All of these results extend the limits of our knowledge in their respective areas and thus shed new light on neutrinos and the physics that surround them.

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-12-01

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 ⁷LiF + BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in ⁷Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

Evaluation of performance of select fusion experiments and projected reactors. Final report

International Nuclear Information System (INIS)

Miley, G.H.

1978-10-01

The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters

Reactors Project Delivery: The Value of Experiance

International Nuclear Information System (INIS)

Stosic, V. Zoran

2014-01-01

State of Affairs: Energy Potential and Density versus Environmental Load of different Energy Sources, Development of Fuel into Energy/Electricity Generation, Production Costs of Electricity, Contributions of Nuclear Energy to Security of Energy Supply, Recent Nuclear Development, Public Support growing again. Projects Status: Reactors under Construction, Different Projects Industrial Schemes, Projects Overview. The Value of Experience: Licensing, Standardization on Early Engineering Activities, Supply Chain and Manufacturing of Heavy Components, Installation, Procurement. (author)

Neutron irradiation experiments for fusion reactor materials through JUPITER program

International Nuclear Information System (INIS)

Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

1998-01-01

A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

International Nuclear Information System (INIS)

1975-12-01

The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

International Nuclear Information System (INIS)

Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

1981-01-01

This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

Experience of on-site disposal of production uranium-graphite nuclear reactor.

Science.gov (United States)

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

Packaging and shipment of U.S. breeder reactor experiments

International Nuclear Information System (INIS)

Berger, J.D.

1980-01-01

Irradiation testing of fuels and materials in the Fast Test Reactor (FTR) required development of a shipping cask (designated T-3) and associated hardware for loading and shipping of these experiments to postirradiation examination facilities. The T-3 shipping-cask program included design, fabrication, and testing of internal cask packages to protect the experiments during loading, shipping, and unloading. The cask was designed for loading in both the vertical and horizontal attitudes

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Operating experience feedback report: Experience with pump seals installed in reactor coolant pumps manufactured by Byron Jackson

International Nuclear Information System (INIS)

Bell, L.G.; O'Reilly, P.D.

1992-09-01

This report examines the reactor coolant pump (RCP) seal operating experience through August 1990 at plants with Byron Jackson (B-J) RCPs. ne operating experience examined in this analysis included a review of the practice of continuing operation with a degraded seal. Plants with B-J RCPs that have had relatively good experience with their RCP seals attribute this success to a combination of different factors, including: enhanced seal QA efforts, modified/new seal designs, improved maintenance procedures and training, attention to detail, improved seal operating procedures, knowledgeable personnel involved in seal maintenance and operation, reduction in frequency of transients that stress the seals, seal handling and installation equipment designed to the appropriate precision, and maintenance of a clean seal cooling water system. As more plants have implemented corrective measures such as these, the number of B-J RCP seal failures experienced has tended to decrease. This study included a review of the practice of continued operation with a degraded seal in the case of PWR plants with Byron Jackson reactor coolant pumps. Specific factors were identified which should be addressed in order to safety manage operation of a reactor coolant pump with indications of a degrading seal

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

International Nuclear Information System (INIS)

Naff, S.A.; Schwarz, W.F.

1978-01-01

Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

Fission product behavior in the Molten Salt Reactor Experiment

International Nuclear Information System (INIS)

Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

1975-10-01

Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

International Nuclear Information System (INIS)

2017-01-01

The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

Energy Technology Data Exchange (ETDEWEB)

Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

1974-07-01

Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

Further analysis of the zero-energy experiment on the Dragon reactor

International Nuclear Information System (INIS)

Woloch, F.; Neuberger, W.

1978-01-01

The analysis of the Zero-Energy Experiments performed on the Dragon reactor, a high-temperature reactor of the Organization for Economic Cooperation and Development, has been continued. The first analysis established the main route of calculations within the WIMS-E scheme and was reported elsewhere. This Note presents further calculations showing the merits of a refinement in the number of neutron energy groups, of the use of different condensation spectra, and of transport calculations

Experiments in power distribution control on the IRT-2000 reactor

International Nuclear Information System (INIS)

Filipchuk, E.V.; Potapenko, P.T.; Trofimov, A.P.; Kosilov, A.N.; Neboyan, V.T.; Timokhin, E.S.

1975-01-01

The results from the experimental investigations of a system for regulating the neutron field on a research reactor IRT-2000 are shown. The right of such experiments on a reactor with a little active zone is substantiated. A successful attempt was made in this work to apply primary elements of straight charging in the neutron field regulating system. A system with independent instrumentally local regulators, a system with hard cross connections and a structure with a ''floating'' installation are studied. Serial common industrial regulators BRT-2 were used

Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

Energy Technology Data Exchange (ETDEWEB)

Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1995-09-01

There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

Energy Technology Data Exchange (ETDEWEB)

Wood, RT

2004-09-27

This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

Experiment calculated ascertainment of factors affecting the energy release in IGR reactor core

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Zhotabayev, Zh.R.

2006-01-01

Full text: At present energy supply resources problem is important. Nuclear reactors can, of course, solve this problem, but at the same time there is another issue, concerning safety exploitation of nuclear reactors. That is why, for the last seven years, such experiments as 'Investigation of the processes, conducting severe accidents with core melting' are being carried out at our IGR (impulse graphite reactor) reactor. Leaving out other difficulties of such experiments, it is necessary to notice, that such experiments require more accurate IGR core energy release calculations. The final aim of the present research is verification and correction of the existing method or creation of new method of IGR core energy release calculation. IGR reactor is unique and there is no the same reactor in the world. Therefore, application of the other research reactor methods here is quite useful. This work is based on evaluation of factors affecting core energy release (physical weight of experimental device, different configuration of reactor core, i.e. location of absorbers, initial temperature of core, etc), as well as interference of absorbers group. As it is known, energy release is a value of integral reactor power. During experiments with rays, Reactor power depends on currents of ion production chambers (IPC), located round the core. It is worth to notice that each ion production chamber (IPC) in the same start-up has its own ratio coefficient between IPC current and reactor present power. This task is complicated due to 'IPC current - reactor power' ratio coefficients, that change continuously, probably, because of new loading of experimental facility and different position of control rods. That is why, in order to try about reactor power, before every start-up, we have to re-determine the 'IPC current - reactor power' ratio coefficients for each ion production chamber (IPC). Therefore, the present work will investigate the behavior of ratio coefficient within the

Light water reactor mixed-oxide fuel irradiation experiment

International Nuclear Information System (INIS)

Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

1998-01-01

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

Complementarity of integral and differential experiments for reactor physics purposes

International Nuclear Information System (INIS)

Tellier, Henry.

1981-04-01

In this paper, the following topics are studied: uranium 238 effective integral; thermal range uranium 238 capture cross section; Americium 242 m capture cross section. The mentioned examples show that differential and integral experiments are both useful to the reactor physicists

Operating experience feedback report: Progress in scram reduction: Commercial power reactors

International Nuclear Information System (INIS)

Bell, L.G.; O'Reilly, P.D.

1989-03-01

This report documents the results of a trends and patterns analysis of unplanned reactor scrams at commercial US nuclear power reactors from January 1, 1984 to January 1, 1988. Major objectives of this report prepared by the Nuclear Regulatory Commission's (NRC's) Office for Analysis and Evaluation of Operational Data (AEOD) are to: (1) provide feedback of operational experience regarding reactor scram trends in support of the Commission's Strategic Goals, (2) examine the causes of unplanned scrams, and (3) examine the relationship between the causes of unplanned scrams and industry initiatives undertaken to reduce the frequency of unplanned scrams, especially with a view to the potential for future scram rate reduction. 31 refs., 14 figs., 49 tabs

Fast reactor sodium systems operation experience and 'leak-before-break' criterion

International Nuclear Information System (INIS)

Ivanenko, V.N.; Zybin, V.A.

1996-01-01

In the paper sodium leakage detection systems used at fast reactors are described. Requirements on their main characteristics (sensitivity, response lime) are formulated. Results of tests are presented on studying the parameters of sodium leak detection systems including experiments on the measurement of size distribution of aerosol particles that have passed through sodium systems thermal insulation after leak initiation. Comparison of these data with dispersion of particles formed at free burning is carried out. Experience of real leaks that occurred at fast reactor sodium systems is analyzed. It has been shown that initiation and development of real leaks do not always follow the theoretical scheme. A substantial role of human factor for sodium systems reliability relative to sodium leaks is stressed. (author)

Vibration behavior of PWR reactor internals Model experiments and analysis

International Nuclear Information System (INIS)

Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

1975-01-01

In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

Energy Technology Data Exchange (ETDEWEB)

Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

2015-09-01

The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

China: EDF's feedback experience of reactor operating is essential to win international markets

International Nuclear Information System (INIS)

Maillart, H.

2016-01-01

The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

Experiment monitoring system of a new electromagnet drive for nuclear reactor control rod

International Nuclear Information System (INIS)

Zhang Jige; Wang Xiaoguang; Wu Yuanqiang; Zhang Zhengming

2003-01-01

In order to deal with some unsolved problems in the engineering prototype design of a new electromagnet drive device for nuclear reactor control rod, the property experiment in view of principle prototype is carried out. Actual displacement of nuclear reactor control rod is measured by means of raster ruler and the test data is obtained by means of computer. The computer communicates with PLC using RS232 serial port. The experimental results show that the monitoring system have the properties of high reliability and high precision, and ensures the experiment to accomplish successfully

Mass hierarchy sensitivity of medium baseline reactor neutrino experiments with multiple detectors

Energy Technology Data Exchange (ETDEWEB)

Wang, Hong-Xin, E-mail: hxwang@iphy.me [Department of Physics, Nanjing University, Nanjing 210093 (China); Zhan, Liang; Li, Yu-Feng; Cao, Guo-Fu [Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049 (China); Chen, Shen-Jian [Department of Physics, Nanjing University, Nanjing 210093 (China)

2017-05-15

We report the neutrino mass hierarchy (MH) determination of medium baseline reactor neutrino experiments with multiple detectors, where the sensitivity of measuring the MH can be significantly improved by adding a near detector. Then the impact of the baseline and target mass of the near detector on the combined MH sensitivity has been studied thoroughly. The optimal selections of the baseline and target mass of the near detector are ∼12.5 km and ∼4 kton respectively for a far detector with the target mass of 20 kton and the baseline of 52.5 km. As typical examples of future medium baseline reactor neutrino experiments, the optimal location and target mass of the near detector are selected for the specific configurations of JUNO and RENO-50. Finally, we discuss distinct effects of the reactor antineutrino energy spectrum uncertainty for setups of a single detector and double detectors, which indicate that the spectrum uncertainty can be well constrained in the presence of the near detector.

Development of the “Approach to Critical” Experiment Simulation Model for the Consort Reactor Using LABVIEW

International Nuclear Information System (INIS)

Abbo, D. N. O.

2015-01-01

Following the shutdown of the CONSORT reactor, the “Approach to Critical” experiment which allowed students to observe and understand the procedure for taking the reactor to critical, balancing the system at low power and increasing the power over a range of powers levels and eventual reactor shutdown, would no longer be possible. It was therefore important to develop a simulation model of the experiment that would enable future students to have comparable training. An “Approach to Critical” Experiment Simulation model for the CONSORT Reactor was developed using Lab-VIEW software to simulate the “Students” experiment version. Lab– VIEW software was chosen due to its good user graphical user interface, offers ready to start functions and also the possibility of improving on the system with new algorithms. The modulation process was used to develop mathematical codes from equations using Lab–VIEW 2012 based on the CONSORT historical experimental data and known literature. The Simulation models the kinetics of a sub-critical reactor with a start-up neutron source, such that control rods are used to increase the power, then achieve power balance and finally shutting down the reactor. Reactivity changes due to temperature effects were neglected. The model was validated by testing the code through performing the three parts of the experiment; Approach to Critical, Doubling time method and Rod drop method, and results compared to the historical experimental data. The results were in agreement with historical data. However the negligible variations were obtained in the Rod drop method due to the reactivity values used to generate the code. (author)

Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

International Nuclear Information System (INIS)

Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

1986-05-01

In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

Basic experiments of reactor physics using the critical assembly TCA

International Nuclear Information System (INIS)

Obara, Toru; Igashira, Masayuki; Sekimoto, Hiroshi; Nakajima, Ken; Suzaki, Takenori.

1994-02-01

This report is based on lectures given to graduate students of Tokyo Institute of Technology. It covers educational experiments conducted with the Tank-Type Critical Assembly (TCA) at Japan Atomic Energy Research Institute in July, 1993. During this period, the following basic experiments on reactor physics were performed: (1) Critical approach experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, (5) Measurement of safety sheet worth by the rod drop method. The principle of experiments, experimental procedure, and analysis of results are described in this report. (author)

Experience with reactor assembly of FBTR

International Nuclear Information System (INIS)

Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

2006-01-01

Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

International Nuclear Information System (INIS)

Cacciapouti, R.J.; Petrusha, L.

1994-01-01

Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented

The RERTR demonstration experiments program at the Ford Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

1983-08-01

The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

International Nuclear Information System (INIS)

Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

1982-01-01

Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

NEA activities in preserving, evaluating and applying data from fast reactor experiments

International Nuclear Information System (INIS)

Gulliford, Jim; Cornet, S.M.; Hill, I.; Yamaji, A.

2013-01-01

Conclusions: Progress to date: • Extensive programme of work to preserve and evaluate data from integral experiments has been established since the mid 1990s. • NEA Data Bank maintains and distributes several databases of these integral experiments, notably through the ICSBEP and IRPhE projects. • More recently programmes of work have been established to help preserve data from the UK Fast Reactor Programme and from various experiments related to minor actinide management. • Data obtained from these programmes are made available to the nuclear science community to provide high quality benchmarks against which modelling methods can be validated. • Involvement of younger scientists and engineers to work alongside well-established experts in the process of evaluating the information is a highly efficient means of transmitting tacit knowledge to the new generation of nuclear specialists. Conclusions: Looking ahead - • Further development of Databases and Database tools, e.g. – improved coverage of fast reactor experiments, MAs; – improved treatment of correlations in uncertainties between experiments; – production of sensitivities to facilitate identification of similar experiments. • Continuation securing UK archives and creating framework for information: – Start identifying suitable integral experiments for inclusion in NEA databases

Proceedings of US-Japan workshop on new generation experiments and reactors (joined by EC)

International Nuclear Information System (INIS)

1988-07-01

The workshop, titled 'New Generation Experiments and Reactors', was held at Plasma Physics Laboratory, Kyoto University from 25 to 28 July 1988. The purpose of the meeting was to review the latest achievements and status of stellarator/heliotron new generation experiments as well as the prospects for stellarator/heliotron fusion reactors on the occasion when the New Large Helical System of MOE in Japan is being realized. The reports on the New Large Helical System of MOE cover an overview, physics issues, design, MHD studies, transport code results and bootstrap current, particle orbit studies, divertor studies, NBI heating, analysis of wave heating, heating system, diagnostics, and SC coil technology. The reports on ATF II cover an overview, physics studies strategy, status of physics studies, engineering issues, perspective of helical systems, issues for next-generation experiments and relationship to the Univ. of Wisconsin Program, and issues for next-generation experiments and relation to Auburn Program. Other reports address recent studies of present devices, studies related with WVIIX, and reactor studies. (N.K.)

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

1977-01-01

The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

Some scoping experiments for a space reactor

International Nuclear Information System (INIS)

Alexander, C.A.; Ogden, J.S.

1983-01-01

Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO 2 and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor

Trends and experiences in reactor coolant pump motors

International Nuclear Information System (INIS)

Anon.

1980-01-01

A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

International Nuclear Information System (INIS)

Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

1982-01-01

During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

The radon monitoring system in Daya Bay Reactor Neutrino Experiment

International Nuclear Information System (INIS)

Chu, M.C.; Kwan, K.K.; Kwok, M.W.; Kwok, T.; Leung, J.K.C.; Leung, K.Y.; Lin, Y.C.; Luk, K.B.; Pun, C.S.J.

2016-01-01

We developed a highly sensitive, reliable and portable automatic system (H 3 ) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H 3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m 3 . This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

Field experience in use of radiation instruments in Cirus reactor

International Nuclear Information System (INIS)

Ramesh, N.; Sharma, R.C.; Agarwal, S.K.; Sawant, D.K.; Yadav, R.K.B.; Prasad, S.K.

2005-01-01

Cirus, located at Bhabha Atomic Research Centre, is a 40 MW (Th) research reactor fuelled by natural uranium, moderated by heavy water and cooled by de-mineralized light water. Graphite is used as reflector in this reactor. The reactor, commissioned in the year 1960, was in operation with availability factor of about 70% till early nineties. There after signs of ageing started surfacing up. After ageing studies, refurbishment plan was finalized and executed during the period from 1997-2002. after successful refurbishment, the reactor is in operation at full power. A wide range of radiation instruments have been used at Cirus for online monitoring of the radiological status of various process systems and environmental releases. Also, variety of survey meters, counting systems and monitors have been used by the health physics unit of the reactor for radiation hazard control. Many of these instruments, which were originally of Canadian design, have undergone changes due to obsolescence or as part of upgradation. This paper describes the experience with the radiation instruments of Cirus, bringing out their effectiveness in meeting the design intent, difficulties faced in their field use, and modifications carried out based on the performance feed back. Also, this paper highlights the areas where further efforts are needed to develop nuclear instrumentation to further strengthen monitoring and surveillance. (author)

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

International Nuclear Information System (INIS)

Tomberlin, T.A.

2002-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed

Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

1994-01-01

Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Yule, T.J.; Fellhauer, C.R.; Boing, L.E.

2002-01-01

A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

Operating experience and maintenance at the TRIGA Mark II LENA reactor

International Nuclear Information System (INIS)

Cingoli, F.; Altieri, S.; Lana, F.; Rosti, G.; Alloni, L.; Meloni, S.

1988-01-01

The last two years at the Trigs Mark II LENA plant were characterized by the running of the n-n-bar oscillation NADIR experiment. Consequently reactor operation was positively affected and the running hours rose again above 1000 hours per year. The LENA team was also deeply involved in the procedures for the renewal of the reactor operation license. The new requirements set by the Nuclear Energy Licensing Authority (ENEA for Italy) most of which concerning radiation protection and environmental impact, have been already fulfilled. In some cases the installation of new apparatus is underway

Development, operational experience and implications for future design of fast reactors in Western Europe

International Nuclear Information System (INIS)

Brandstetter, A.; Broomfield, A.M.; Saitcevsky, B.

1990-01-01

Over the past 30 years, the partners now collaborating in Europe on fast-reactor development have taken the technology from a theoretical possibility to an engineering reality. In that time there has been a progression from experimental zero-energy facilities followed by small power-producing engineering test reactors, to prototype reactors and a large commercial-size reactor. The paper describes the highlights of the reactor programmes in the partner countries and by example illustrates the experience gained from reactor operation and some of the principal activities in the supporting development programme. These include such topics as fuel performance, fast-neutron physics, liquid-metal thermal hydraulics, sodium chemistry, instrumentation and safety aspects. The paper concludes by summarizing some of the main objectives of the current development programme, which is directed to the support of the European Fast Reactor design being prepared by the European design and construction companies. (author)

Fast reactor operating experience gained in Russia: Analysis of anomalies and abnormal operation cases

International Nuclear Information System (INIS)

Ashurko, Y.M.; Baklushin, R.P.; Zagorulko, Y.I.; Ivanenko, V.N.; Matveyev, V.P.; Vasilyev, B.A.

2000-01-01

Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

Educational reactor-physics experiments with the critical assemble TCA

Energy Technology Data Exchange (ETDEWEB)

Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

1997-10-01

The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

Educational reactor-physics experiments with the critical assembly TCA

International Nuclear Information System (INIS)

Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki; Horiki, Oichiro; Suzaki, Takenori.

1997-10-01

The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for 1) Critical approach and Exponential experiment, 2) Measurement of neutron flux distribution, 3) Measurement of power distribution, 4) Measurement of fuel rod worth distribution, and 5) Measurement of safety plate worth by the rod drop method. (author)

HEU to LEU conversion experience at the UMass-Lowell research reactor

International Nuclear Information System (INIS)

White, John R.; Bobek, Leo M.

2005-01-01

The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Tigliole, A. Borio Di; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Provasi, M.C.; Salvini, A.; Scian, G.; Vinciguerra, G. [University of Pavia, Laboratory of Applied Nuclear Energy (L.E.N.A), Via Aselli 41, 27100 Pavia (Italy)

2011-07-01

The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

Benefits of reactor physics experiments for the HTGR industrial development - an attempt to a quantitative approach

Energy Technology Data Exchange (ETDEWEB)

Cuniberti, R; Graziani, G; Massino, L; Rinaldini, C; Zanantoni, C

1972-10-15

The available results of reactor physics experiments on HTGRs and their accuracies are briefiy reviewed. The physical quantities of interest are grouped into three categories: basic nuclear data, lattice parameters and integral design data. The last two are considered and their possible improvements in accuracy by means of experimental measurements are assessed. The cost penalty on fuel cycle and capital cost due to each physical quantity is then considered, and consequently the benefits of reactor physics experiments are evaluated for a number of hypotheses concerning the foreseeable HTGR development and the delay in taking practical advantage of experimental results. It is concluded that, at the present state of knowledge of basic nuclear data and with the available calculation methods, the economic incentive to new reactor physics experiments is small, and a previous careful analysis is recommended to those intending to perform such experiments.

DABIE: a data banking system of integral experiments for reactor core characteristics computer codes

International Nuclear Information System (INIS)

Matsumoto, Kiyoshi; Naito, Yoshitaka; Ohkubo, Shuji; Aoyanagi, Hideo.

1987-05-01

A data banking system of integral experiments for reactor core characteristics computer codes, DABIE, has been developed to lighten the burden on searching so many documents to obtain experiment data required for verification of reactor core characteristics computer code. This data banking system, DABIE, has capabilities of systematic classification, registration and easy retrieval of experiment data. DABIE consists of data bank and supporting programs. Supporting programs are data registration program, data reference program and maintenance program. The system is designed so that user can easily register information of experiment systems including figures as well as geometry data and measured data or obtain those data through TSS terminal interactively. This manual describes the system structure, how-to-use and sample uses of this code system. (author)

Mass hierarchy sensitivity of medium baseline reactor neutrino experiments with multiple detectors

Directory of Open Access Journals (Sweden)

Hong-Xin Wang

2017-05-01

Full Text Available We report the neutrino mass hierarchy (MH determination of medium baseline reactor neutrino experiments with multiple detectors, where the sensitivity of measuring the MH can be significantly improved by adding a near detector. Then the impact of the baseline and target mass of the near detector on the combined MH sensitivity has been studied thoroughly. The optimal selections of the baseline and target mass of the near detector are ∼12.5 km and ∼4 kton respectively for a far detector with the target mass of 20 kton and the baseline of 52.5 km. As typical examples of future medium baseline reactor neutrino experiments, the optimal location and target mass of the near detector are selected for the specific configurations of JUNO and RENO-50. Finally, we discuss distinct effects of the reactor antineutrino energy spectrum uncertainty for setups of a single detector and double detectors, which indicate that the spectrum uncertainty can be well constrained in the presence of the near detector.

Status of the Daya Bay Reactor Neutrino Oscillation Experiment

International Nuclear Information System (INIS)

Lin, Cheng-Ju Stephen

2010-01-01

The last unknown neutrino mixing angle θ 13 is one of the fundamental parameters of nature; it is also a crucial parameter for determining the sensitivity of future long-baseline experiments aimed to study CP violation in the neutrino sector. Daya Bay is a reactor neutrino oscillation experiment designed to achieve a sensitivity on the value of sin 2 (2*θ 13 ) to better than 0.01 at 90% CL. The experiment consists of multiple identical detectors placed underground at different baselines to minimize systematic errors and suppress cosmogenic backgrounds. With the baseline design, the expected anti-neutrino signal at the far site is about 360 events per day and at each of the near sites is about 1500 events per day. An overview and current status of the experiment will be presented.

Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

International Nuclear Information System (INIS)

Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

1983-02-01

This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

Implications of recent implantation-driven permeation experiments for fusion reactor safety

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Struttmann, D.A.

1986-01-01

Metal structures exposed to the plasma in tritium-burning fusion reactors will be subject to implantation-driven permeation (IDP) of tritium. Permeation rates for IDP in fusion structural materials are usually high because the tritium atoms enter the material without having to go through the dissociation and solution steps required of tritium-bearing gas molecules. These surface processes, which may be rate limiting in PDP, actually enhance permeation in IDP by inhibiting the return of tritium to the plasma side of the structure. Experiments have been conducted at the Idaho National Engineering Laboratory (INEL) to investigate the nature of IDP by simulating conditions experienced by structures exposed to the plasma. These experiments have shown that surface conditions are important to tritium permeation in materials endothermic to hydrogen solution such as austenitic and ferritic steels. In reactive metals such as vanadium, surface processes appear to totally control the permeation. The purpose of this paper is to review the progress of those experiments and to discuss the implications that the results have regarding the tritium-related safety concerns of fusion reactors

Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

International Nuclear Information System (INIS)

Cave, L.; Clarke, A.W.

1984-01-01

The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

Irradiation Experiments on Plutonium Fuels for Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

1967-09-15

An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

International Nuclear Information System (INIS)

Wood, RT

2004-01-01

This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

NEA Activities in Preserving, Evaluating and Applying Data from Fast Reactor Experiments

International Nuclear Information System (INIS)

Gulliford, N.T.; Cornet, S.M.; Hill, I.; Yamaji, A.

2015-01-01

The goal of the OECD Nuclear Energy Agency (NEA) in the area of nuclear science is to help member countries identify, collate, develop and disseminate the basic scientific and technical knowledge required to ensure safe and reliable operation of current nuclear systems and to develop next generation technologies. Within these general goals, the current nuclear science programme has three key objectives: (i) to help advance the existing scientific knowledge needed to enhance the performance and safety of current nuclear systems, (ii) to contribute to building a solid scientific and technical basis for the development of future generation nuclear systems and (iii) to support the preservation of essential knowledge in the field of nuclear science. As part of the second and third of these objectives, an extensive programme of work to preserve and evaluate data from integral experiments has been established, including reactor physics, shielding and criticality safety experiments on fast reactor systems. Data from experimental facilities are reviewed and, if necessary, archives of information are made safe. This may typically involve the indexing and scanning of key documents and archiving of logbooks, for example. Selected experiments go through a detailed evaluation process and where deemed appropriate, a benchmark description is created in a standardized format for inclusion in one of the NEA Data Bank international databases. This information is used extensively by the international nuclear science community to validate their modelling and simulation tools. The process can be viewed as part of a broader knowledge management function, where information is gathered, evaluated, linked and made accessible to a wide range of users. The presentation gives details of the main databases maintained and developed by the NEA, focusing on those related to fast reactor applications. The status of recent preservation activities for fast reactor archives in the United Kingdom is

Computer network that assists in the planning, execution and evaluation of in-reactor experiments

International Nuclear Information System (INIS)

Bauer, T.H.; Froehle, P.H.; August, C.; Baldwin, R.D.; Johanson, E.W.; Kraimer, M.R.; Simms, R.; Klickman, A.E.

1985-01-01

For over 20 years complex, in-reactor experiments have been performed at Argonne National Laboratory (ANL) to investigate the performance of nuclear reactor fuel and to support the development of large computer codes that address questions of reactor safety in full-scale plants. Not only are computer codes an important end-product of the research, but computer analysis is also involved intimately at most stages of experiment planning, data reduction, and evaluation. For instance, many experiments are of sufficiently long duration or, if they are of brief duration, occur in such a purposeful sequence that need for speedy availability of on-line data is paramount. This is made possible most efficiently by computer assisted displays and evaluation. A purposeful linking of main-frame, mini, and micro computers has been effected over the past eight years which greatly enhances the speed with which experimental data are reduced to useful forms and applied to the relevant technological issues. This greater efficiency in data management led also to improvements in the planning and execution of subsequent experiments. Raw data from experiments performed at INEL is stored directly on disk and tape with the aid of minicomputers. Either during or shortly after an experiment, data may be transferred, via a direct link, to the Illinois offices of ANL where the data base is stored on a minicomputer system. This Idaho-to-Illinois link has both enhanced experiment performance and allowed rapid dissemination of results

Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

International Nuclear Information System (INIS)

Irianto, Djoko; Kanji, T.; Kukita, Y.

1996-01-01

An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

T. A. Tomberlin; S. B. Grover

2004-11-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

T. A. Tomberlin; S. B. Grover

2004-01-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility

International Nuclear Information System (INIS)

Dietze, Klaus; Klippel, Henk Th.; Koning, Arjan; Jacqmin, Robert

2003-01-01

1 - Description: The STEK-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactors STEK (ECN Petten/ Netherlands) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in RCN-209, ECN-10 The 5 STEK configurations cover a broad energy range due to their increasing softness. The experiments are very valuable because of the extensive program of sample reactivity measurements with many fission product nuclides important in reactor burn-up calculations. At first, analyses of the experiments have been performed in Petten. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and JENDL-3.2) are compared in JEF/DOC-861. It contains the following documents: 31 Reports, 2 publications, 5 JEF documents, 4 conferences. 2 - Related or auxiliary programs

Experiment study on thermal mixing performance of HTR-PM reactor outlet

Energy Technology Data Exchange (ETDEWEB)

Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2016-09-15

A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

14th Biennial conference on reactor operating experience plant operations: The human element

International Nuclear Information System (INIS)

Anon.

1989-01-01

Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

Advanced Reactor Fuels Irradiation Experiment Design Objectives

Energy Technology Data Exchange (ETDEWEB)

Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

Advanced Reactor Fuels Irradiation Experiment Design Objectives

International Nuclear Information System (INIS)

Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

2016-01-01

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

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

GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; John D. Bess; Jim Gulliford

2011-09-01

Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

Background determination for the neutron-neutron scattering experiment at the reactor YAGUAR

Energy Technology Data Exchange (ETDEWEB)

Muzichka, A.Yu. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Furman, W.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Krylov, A.R. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V.N. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Levakov, B.G. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Lyzhin, A.E. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Chernukhin, Yu.I. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Kandiev, Ya.Z. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Howell, C.R. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G.E. [North Carolina State University, Raleigh, NC 27695-8202 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Crawford, B.E. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S.L. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States)]. E-mail: sstephen@gettysburg.edu; Tornow, W. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)

2007-06-01

The motivation and design is outlined for the experiment to measure the neutron-neutron singlet scattering length directly with thermal neutrons at the pulsed reactor YAGUAR. A statistical accuracy of 3% can be reached, though achieving the goal of an overall accuracy of 3-5% for the nn-scattering length depends on the background level. Possible sources of background are discussed in depth and the results of extensive modeling of the background are presented. Measurements performed at YAGUAR to test these background calculations are described. The experimental results indicate an anticipated background level up to 30% relative to the expected nn effect at the maximal energy burst of the reactor. The conclusion is made that the nn experiment at YAGUAR is feasible to produce the first directly measured value for the neutron-neutron scattering length.

MHTGR: New production reactor summary of experience base

International Nuclear Information System (INIS)

1988-03-01

Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs

Integral activation experiment of fusion reactor materials with d-Li neutrons up to 55 MeV

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Moellendorff, Ulrich von [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Wada, Masayuki [Business Automation Co., Ltd., Tokyo (Japan)

2000-03-01

An integral activation experiment of fusion reactor materials with a deuteron-lithium neutron source was performed. Since the maximum energy of neutrons produced was 55 MeV, the experiment with associated analysis was one of the first attempts for extending the energy range beyond 20 MeV. The following keywords represent the present study: d-Li neutrons, 55 MeV, dosimetry, SAND-II, spectrum adjustment, LA-150, MCNP, McDeLi, IFMIF, fusion reactor materials, integral activation experiment, low-activation, F82H, vanadium-alloy, IEAF, ALARA, and sequential charged particle reaction. (author)

Chemistry in water reactors: operating experience and new developments. 2 volumes

International Nuclear Information System (INIS)

1994-01-01

These proceedings of the International conference on chemistry in water reactors (Operating experience and new developments), Volume 1, are divided into 8 sessions bearing on: (session 1) Primary coolant activity, corrosion products (5 conferences), (session 2) Dose reduction (4 conferences), (session 3) New developments (4 conferences), poster session: Primary coolant chemistry (16 posters), (session 4) Decontamination (5 conferences), poster session (2 posters), (session 5) BWR-Operating experience (3 conferences), (session 6) BWR-Modelling of operating experience (4 conferences), (session 7) BWR-Basic studies (4 conferences), (session 8) BWR-New technologies (3 conferences)

Recent performance experience with US light water reactor self-actuating safety and relief valves

Energy Technology Data Exchange (ETDEWEB)

Hammer, C.G.

1996-12-01

Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

Experience of developing the imitators of the fuel element for the WWER reactors

International Nuclear Information System (INIS)

Balashov, S.M.; Boltenko, Eh.A.; Vinogradov, V.A.

1998-01-01

Peculiarities of designs of fuel elements imitators for the WWER-type reactors of nominal capacity and with single-ended current feed positioning are considered. The data on the filler heat conductivity and the results of tests and application of the fuel elements imitators at various testing facilities are presented. The possibility of equipping one of the non operating WWER reactors with the fuel element imitators for conduct of large-scale experiment is indicated

Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

International Nuclear Information System (INIS)

Halvers, L.J.

1977-03-01

An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

International Nuclear Information System (INIS)

Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

1979-01-01

In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

International Nuclear Information System (INIS)

Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

1982-01-01

Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components

Reactor design, cold-model experiment and CFD modeling for chemical looping combustion

Energy Technology Data Exchange (ETDEWEB)

Zhang, Shaohua; Ma, Jinchen; Hu, Xintao; Zhao, Haibo; Wang, Baowen; Zheng, Chuguang [Huazhong Univ. of Science and Technology, Wuhan (China). State Key Lab. of Coal Combustion

2013-07-01

Chemical looping combustion (CLC) is an efficient, clean and cheap technology for CO{sub 2} capture, and an interconnected fluidized bed is more appropriate solution for CLC. This paper aims to design a reactor system for CLC, carry out cold-model experiment of the system, and model fuel reactor using commercial CFD software. As for the CLC system, the air reactor (AR) is designed as a fast fluidized bed while the fuel reactor (FR) is a bubbling bed; a cyclone is used for solid separation of the AR exit flow. The AR and FR are separated by two U-type loop seals to remain gas sealed. Considered the chemical kinetics of oxygen carrier, fluid dynamics, pressure balance and mass balance of the system simultaneously, some key design parameters of a CH{sub 4}-fueled and Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3}-based CLC reactor (thermal power of 50 kWth) are determined, including key geometric parameters (reactor cross-sectional area and reactor height) and operation parameters (bed material quantity, solid circulation rate, apparent gas velocity of each reactor). A cold-model bench having same geometric parameters with its prototype is built up to study the effects of various operation conditions (including gas velocity in the reactors and loop seals, and bed material height, etc.) on the solids circulation rate, gas leakage, and pressure balance. It is witnessed the cold-model system is able to meet special requirements for CLC system such as gas sealing between AR and FR, the circulation rate and particles residence time. Furthermore, the thermal FR reactor with oxygen carrier of Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} and fuel of CH{sub 4} is simulated by commercial CFD solver FLUENT. It is found that for the design case the combustion efficiency of CH{sub 4} reaches 88.2%. A few part of methane is unburned due to fast, large bubbles rising through the reactor.

Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

International Nuclear Information System (INIS)

Bagley, K.Q.

1978-01-01

The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

Nuclear engineering laboratory self regulated power oscillation experiments at the Health Physics Research Reactor

International Nuclear Information System (INIS)

Miller, L.F.; Mihalczo, J.T.; Bailiff, E.G.; Woody, N.D.; Gardner, G.D.

1983-01-01

Self regulated power oscillation experiments with a variety of initial conditions have been performed with the ORNL Health Physics Research Reactor (HPRR) by undergraduate nuclear engineering students from The University of Tennessee for several years. These experiments demonstrate the coupling between reactor kinetics and heat transfer and show how the temperature coefficient of reactivity affects reactor behavior. A model that consists of several coupled first order nonlinear differential equations is used to calculate the temperature of the core center and surface and power as a function of time which are compared with the experimental data; also, the model is also used to study the effects of various model parameters and initial conditions on the amplitude, frequency and damping of the power and temperature oscillations. A previous paper presented some limited experimental results and demonstrated the correspondence between a simple point model and the experimental data. This paper presents the results of experiments for: (1) the initial power fixed at 9 kW with central core temperatures of 300 0 F and 500 0 F, annd (2) the initial central core temperature fixed at 500 0 F with initial powers of 6 and 8 kW

Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

International Nuclear Information System (INIS)

Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

1987-09-01

Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

Reactor Neutrino Detection for Non Proliferation with the NUCIFER Experiment

Energy Technology Data Exchange (ETDEWEB)

Bouvet, L. [CEA, Centre de Saclay, IRFU, Gif-sur-Yvette, (France); Bouvier, S.; Bui, V. M. [Laboratoire Subatech, Ecole des Mines, Nantes Cedex 3 (France); others, and

2012-06-15

Neutrinos are the most abundant matter particles in the Universe. Thoroughly investigated in basic science, the neutrino field is now delivering first applications to the monitoring of nuclear reactors. The neutrinos are emitted in the decay chain of the fission products; therefore measuring their flux provides real-time information, directly related to the fission process occurring in the reactor core. Because of the very weak interaction of neutrinos with matter a neutrino detector can stand outside the core containment vessel and provide a non-intrusive and inherently tamper resistant measurement. After a brief review of the existing data and worldwide projects, we present the NUCIFER experiment. The active part of the detector is a tank filled up with one ton of Gadolinium-doped liquid scintillator. Sixteen photomultiplier tubes, isolated from the liquid by an acrylic buffer, read out the light produced by the interaction of a neutrino with the protons of the liquid. The tank is surrounded by plastic scintillator plates to veto the cosmic rays. Then polyethylene and lead shielding suppress the background coming from external neutrons and gamma rays respectively. The NUCIFER detector has been designed for an optimal compromise between the detection performances and the specifications of operation in a safeguards regime. Its global footprint is 2.8 m x 2.8 m and it can monitor remotely the nuclear power plant thermal power and Plutonium content with very little maintenance on years scale. The experiment is currently installed near the OSIRIS research reactor (70 MWth) at CEA, in Saclay, France. First data are expected by May 2012. This work is done in contact with the IAEA/SGTN division that is currently investigating the potentiality of neutrinos as a novel safeguards tool. A dedicated working group has been created in 2010 to coordinate the simulation effort of various reactor types as well as the development of dedicated detectors and define and eventually

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

Energy Technology Data Exchange (ETDEWEB)

Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

2015-03-01

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.; Aguilar H, F.; Rivero G, T.; Sainz M, E.

2000-01-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Evaluation guide for the international reactor physics experiments evaluation project (IRPhEP)

International Nuclear Information System (INIS)

Yamaji, Akifumi

2013-01-01

At present, there is an urgent need to preserve integral reactor physics experimental data including separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. The International Reactor Physics Evaluation Project (IRPhEP) was initiated as a pilot activity in 1999 by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. This document outlines the general presentation guidelines that evaluators should follow for the description of the experiments and all relevant experimental data in order to ensure the consistency between the evaluations published in the final Handbook. Publication templates will be used to ensure this consistency and will follow the general scheme below: 1 - Experiment identification number; 2- Date; 3 - Name of experiment (Purpose of experiment, Phenomena measured and scope); 4 - Name or designation of experimental programme; 5 - Description of facility; 6 - Description of test or experiment (Experimental configuration, Core life cycle, Experimental limitations or shortcomings); 7 - Phenomena measured (Description of results and analysis, Special features and characteristics of experiment, Measurement systems/methods and uncertainties); 8 - Duplicate or complementary experiments / other related experiments; 9 - Status of completion of the evaluation; 10 - References (pointer to evaluation, archive if available, otherwise generic bibliographic reference); 11 - Authors/ organisers 12 - Material available

Design experiences for medical irradiation field at the musashi reactor

International Nuclear Information System (INIS)

Aizawa, Otohiko

1994-01-01

The design of the medical irradiation field at the Musashi reactor was carried out from 1974 to 1975, about 20 years ago. Various numerical analyses have been carried out recently, and it is astonishing to find out that the performance close to the optimum as a 100 kW reactor has been obtained. The reason for this is that the design was carried out by dividing into the stationary part and the moving part, and as for the moving part, the structure was determined by repeating trial and error and experiments. In this paper, the comparison of the analysis carried out later with the experimental data and the change of the absorbed dose at the time of medical irradiation accompanying the change of neutron energy spectra are reported. As the characteristics of the medical irradiation field at the Musashi reactor, the neutron energy spectra and the absorbed dose and mean medical irradiation time are shown. As the problems in boron neutron capture therapy, the neutron fluence required for the therapy, the way of thinking on background dose, and the problem of determining the irradiation time are discussed. The features of epithermal neutron beam are explained. (K.I.)

Meeting the future of coherent neutrino scattering. A feasibility study for upcoming reactor experiments

Energy Technology Data Exchange (ETDEWEB)

Salathe, Marco; Rink, Thomas [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

2016-07-01

Due to ongoing progress in detector development and background suppression techniques first evidence of neutrino coherent scattering seems reachable in future experiments. In recent years efforts have been enhanced to detect this effect with germanium detectors. This work aims at summarizing and improving past studies on the potential of an experiment at a reactor site to a new level of accuracy by using the most recent neutrino spectra, knowledge gained in recent detector developments and in contrast to prior studies an energy-dependent quenching factor. The influence of the main parameters (background suppression, detector resolution and threshold, reactor spectra, different isotopes) of a germanium detector experiment is presented and the sensitivities regarding the main reaction channels are calculated. The results were obtained through two independent methods; an algebraic computation and a numerical simulation. Both methods reveal the most important experimental parameters and clarify the state of the art challenges that research has to meet in such an experiment.

Effect of Drawer Master Modeling of ZPPR15 Phase A Reactor Physics Experiment on Integral Parameter

International Nuclear Information System (INIS)

Yoo, Jae Woon; Kim, Sang Ji

2011-01-01

As a part of an International-Nuclear Engineering Research Initiative (I-NERI) Project, KAERI and ANL are analyzing the ZPPR-15 reactor physics experiments. The ZPPR-15 experiments were carried out in support of the Integral Fast Reactor (IFR) project. Because of lack of the experimental data, verifying and validating the core neutronics analysis code for metal fueled sodium cooled fast reactors (SFR) has been one of the big concerns. KAERI is developing the metal fuel loaded SFR and plans to construct the demonstration SFR by around 2028. Database built through this project and its result of analysis will play an important role in validating the SFR neutronics characteristics. As the first year work of I-NERI project, KAERI analyzed ZPPR-15 Phase A experiment among four phases (Phase A to D). The effect of a drawer master modeling on the integral parameter was investigated. The approximated benchmark configurations for each loading were constructed to be used for validating a deterministic code

Critical experiment and analysis for nitride fuel fast reactor using FCA

International Nuclear Information System (INIS)

Andoh, Masaki; Iijima, Susumu; Okajima, Shigeaki; Sakurai, Takeshi; Oigawa, Hiroyuki

2000-03-01

As a research on FBR with new types of fuel, a series of experiments on a nitride fuel fast reactor was carried out at Fast Critical Assembly (FCA) to evaluate the calculation accuracy on the neutronic characteristics of the reactor. In this study, criticality, sample reactivity worth and sodium void reactivity worth were measured in the FCA XIX-2 core simulating a nitride fuel fast reactor and were analyzed using the standard analysis method for FCA fast reactor cores. The accuracy of the analysis on the effective multiplication factor was the same as those of the other FCA cores. For the plate sample reactivity worth, the calculation on the radial distribution of plutonium plate reactivity worth overestimated the measurement depending on the distance from the center of the core. For the sodium void reactivity worth, the calculation overestimated the experimental value 10 to 20% at the core center, while the overestimation was improved as the voided position was located at the core boundary. It was found that the transport effect was considerable even at the center of the core. It was considered that the calculation accuracy on the non-leakage term of the void reactivity worth and transport correction should be improved. (author)

Split core experiments; Part I. Axial neutron flux distribution measurements in the reactor core with a central horizontal reflector

Energy Technology Data Exchange (ETDEWEB)

Strugar, P; Raisic, N; Obradovic, D; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1965-05-01

A series of critical experiments were performed on the RB reactor in order to determine the thermal neutron flux increase in the central horizontal reflector formed by a split reactor core. The objectives of these experiments were to study the possibilities of improving the thermal neutron flux characteristics of the neutron beam in the horizontal beam tube of the RA research reactor. The construction of RA reactor enables to split the core in two, to form a central horizontal reflector in front of the beam tube. This is achieved by replacing 2% enriched uranium slugs in the fuel channel by dummy aluminium slugs. The purpose of the first series of experiments was to study the gain in thermal neutron component inside the horizontal reflector and the loss of reactivity as a function of the lattice pitch and central reflector thickness.

Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

International Nuclear Information System (INIS)

Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

2010-01-01

The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

Energy Technology Data Exchange (ETDEWEB)

Paar, H; Narodoslawsky, M; Moser, A [Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie

1990-10-10

Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

Proceedings of the CNRA Workshop on New Reactor Siting, Licensing and Construction Experience

International Nuclear Information System (INIS)

2011-01-01

This report documents the proceedings from the 'Workshop on New Reactor Siting, Licensing and Construction Experience', held in Prague, Czech Republic on 15-17 September 2010. A total of 59 specialists from 16 countries and international organisations attended. The Meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the State Office for Nuclear Safety (SUJB) in Czech Republic. The objectives of the workshop were to review and discuss recent and past construction experience lessons learned including perspectives from regulatory authorities, as well as vendors, and licensee. The workshop addressed issues associated with project management resources including: a) overall human resources, expertise, experience and organisation available to the licensee, b) capability of each potential vendor (in-house knowledge and skills versus planned subcontracting and subcontractor management). The workshop also discussed the lessons learned in the regulation of site selection, evaluation and site preparation as well as the review of regulatory practices for the licensing of new reactors, including the regulatory body infrastructure, staffing and expertise needed. The workshop provided an excellent opportunity to communicate recent experience on these topics to a wider audience, including participants from OECD member countries as well as New Entrants from non-OECD member countries. The workshop allowed the WGRNR group to introduce and discuss the current programme of work and products under development in order to gain insights from workshop participants on each of the programme of work areas, and get feedback on additional focus areas. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to presentations of the licensing process for new reactors followed by different member countries. The second technical session was

Summary of the fourth conference on United States utility experience in reactor noise analysis

International Nuclear Information System (INIS)

Fry, D.N.

1987-01-01

The fourth informal conference on United States utility experience in reactor noise analysis and loose-part monitoring was held at the Northeast Utilities Service Company offices in Hartford, Connecticut, May 12-14, 1987. Host and general chairman for the meeting was J.V. Persio of Northeast Utilities. This conference provided a forum where utilities could share information on reactor noise analysis on an informal basis. There were about 60 attendees at the meeting representing 10 utilities, 3 reactor vendors, 8 consulting organizations, and 4 universities and research laboratories. Twenty-three papers were presented at the conference, dealing with various aspects of loose-part monitoring, neutron noise analysis, and standards activities

The operating experience and incident analysis for High Flux Engineering Test Reactor

International Nuclear Information System (INIS)

Zhao Guang

1999-01-01

The paper describes the incidents analysis for High Flux Engineering test reactor (HFETR) and introduces operating experience. Some suggestion have been made to reduce the incidents of HFETR. It is necessary to adopt new improvements which enhance the safety and reliability of operation. (author)

Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

Energy Technology Data Exchange (ETDEWEB)

A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

2011-07-01

High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

Energy Technology Data Exchange (ETDEWEB)

Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

2015-07-01

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

Practical experience in the application of quality control in water-reactor fuel fabrication

International Nuclear Information System (INIS)

Vollath, D.

1984-07-01

Highly industrialized countries have gained vast experience in manufacturing water reactor fuel. Manufacturing is followed by a stringent system of quality assurance and quality control. The Seminar on Practical Experience in the Application of Quality Control in Water-Reactor Fuel Fabrication provided a forum for an exchange of information on methods and systems of quality assurance and quality control for reactor fuel. In addition, many developing countries which have started or intend to set up a nuclear fuel industry are interested in the application of quality assurance and quality control. This meeting has been preceded by two different series of conferences: the IAEA meetings 1976 in Oslo, 1978 in Prague and 1979 in Buenos Aires, and the Karlsruhe meetings on Characterization and Quality Control of Nuclear Fuel held in 1978 and 1981. Quality control and quality assurance has many different facets. Unlike the purely technical aspects, covered by the Karlsruhe conference series, the IAEA meetings always relate to a wider field of topics. They include governmental regulations and codes for practical quality assurance. This volume contains the papers presented at the seminar and a record of the discussions. (orig.)

Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

1997-01-01

For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Operational experience with the TRIGA reactor of the University of Pavia

International Nuclear Information System (INIS)

Borio di Tigliole, A.; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Salvini, A.; Scian, G.; Vinciguerra, G.

2008-01-01

The TRIGA Mark II research reactor of the University of Pavia is in operation since 1965. The annual operational time at nominal power (250 kW) is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities and BNCT research. Few tens of hours per year are dedicated also to electronic devices irradiation and student training courses. Few homemade upgrading of the reactor were realized in the past two years: components of the secondary/tertiary cooling circuit were substituted and a new radiation area monitoring system was installed. Also the Instrumentation and Control (I and C) system was almost completely refurbished. The presentation describes the major extraordinary maintenance activities implemented and the status of main reactor systems: - The I and C System: complete substitution, channel-by-channel without changing the operating and safety logics; - Tertiary and secondary water-cooling circuits: complete substitution of the tertiary water-cooling circuit and partial substitution of the components of the secondary water-cooling circuit; - Reactor Building Air Filtering and Ventilation System: installation of a computerized air filtering and ventilation system; - Radiation Area Monitoring System: new system based on a commercial micro-computer and an home-made software developed on Lab-View platform. The system is made of a network of different instruments coupled, trough a serial bus line RS232, with a data acquisition station; - Fuel Elements: at the moment, the core is made of 48 Aluminium clad and 34 SST clad TRIGA fuel elements controlled periodically for their elongation and/or bowing. All components and systems undergo ordinary maintenance according to the Technical Prescriptions and to the 'Good Practice Procedures'. In summary, the TRIGA reactor of the University of Pavia shows a very good technical state and, at the moment, there are no political or

Analysis of kyoto university reactor physics critical experiments using NCNSRC calculation methodology

International Nuclear Information System (INIS)

Amin, E.; Hathout, A.M.; Shouman, S.

1997-01-01

The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab

System Requirements Document for the Molten Salt Reactor Experiment 233U conversion system

International Nuclear Information System (INIS)

Aigner, R.D.

2000-01-01

The purpose of the conversion process is to convert the 233 U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019

On-site underground background measurements for the KASKA reactor-neutrino experiment

International Nuclear Information System (INIS)

Furuta, H.; Sakuma, K.; Aoki, M.; Fukuda, Y.; Funaki, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Katsumata, M.; Kawasaki, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Matsumoto, T.; Miyata, H.; Nagasaka, Y.; Nakagawa, T.; Nakajima, N.; Nitta, K.; Sakai, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.; Tamura, N.; Tsuchiya, Y.

2006-01-01

On-site underground background measurements were performed for the planned reactor-neutrino oscillation experiment KASKA at Kashiwazaki-Kariwa nuclear power station in Niigata, Japan. A small-diameter boring hole was excavated down to 70m underground level, and a detector unit for γ-ray and cosmic-muon measurements was placed at various depths to take data. The data were analyzed to obtain abundance of natural radioactive elements in the surrounding soil and rates of cosmic muons that penetrate the overburden. The results will be reflected in the design of the KASKA experiment

Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

International Nuclear Information System (INIS)

Hine, R.E.

1980-09-01

This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated

Model experiments on simulation of the WWER water-chemical conditions at loop facilities of the MIR reactor

International Nuclear Information System (INIS)

Benderskaya, O.S.; Zotov, E.A.; Kuprienko, V.A.; Ovchinnikov, V.A.

1999-01-01

The experiments on simulation of the WWER type reactors water-chemical conditions have been started at the State Scientific Center RIAR. These experiments are being conducted at the multi-loop research MIR reactor at the PVK-2 loop facility. The dosage stand was created. It allows introduction of boric acid, potassium and lithium hydroxides, ammonia solutions and gaseous hydrogen. Corrosion tests of the Russian E-635 and E-110 alloys are being conducted at the PVK-2 loop under the WWER water-chemical conditions. If necessary, fuel elements are periodically extracted from the reactor to perform visual examination, to measure their length, diameter, to remove the deposits from the claddings, to measure the burnup and to distribute the fission products over the fuel element by gamma-spectrometry. The chemical analytical 'on line' equipment produced by the ORBISPHERE Laboratory (Switzerland) will be commissioned in the nearest future to measure concentration of the dissolved hydrogen and oxygen as well as pH and specific conductivity. The objective of the report is to familiarize the participants of the IAEA Technical Committee with the capabilities of performing the model water-chemical experiments under the MIR reactor loop facility conditions. (author)

A boiling-water reactor concept for low radiation exposure based on operating experience

International Nuclear Information System (INIS)

Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

1983-01-01

A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (author)

TSUNAMI analysis of the applicability of proposed experiments to reactor-grade and weapons-grade mixed-oxide systems

International Nuclear Information System (INIS)

Rearden, Bradley T.; Hopper, Calvin M.; Elam, Karla R.

2005-01-01

The applicability of proposed critical experiments for the criticality code validation of a series of prototypic reactor-grade and weapons-grade mixed-oxide systems has been assessed with the TSUNAMI methodology from SCALE 5. The application systems were proposed by the Nuclear Energy Agency (NEA) Organization for Economic Cooperation and Development (OECD) Working Party on Nuclear Criticality Safety MOX Experimental Needs Working Group. Forty-eight application systems were conceived to envelope the range of conditions in processing and fabrication of reactor-grade and weapons-grade MOX fuel. The applicability of 303 existing critical benchmarks to each of the 48 applications was assessed, and validation coverage was found to be lacking for certain applications. Two series of proposed critical experiments were also considered in this analysis. The TSUNAMI analysis has revealed that both series of proposed experiments are applicable to numerous configurations of the reactor-grade and weapons-grade systems. A detailed assessment of which experiments were revealed by TSUNAMI to be most applicable to specific prototypic fuel processing systems has been performed. (author)

Analysis of the Rossendorf SEG experiments using the JNC route for reactor calculation

International Nuclear Information System (INIS)

Dietze, Klaus

1999-11-01

The integral experiments performed at the Rossendorf fast-thermal coupled reactor RRR/SEG have been reanalyzed using the JNC route for reactor calculation JENDL3.2/SLAROM/CITATION/JOINT/PERKY. The Rossendorf experiments comprise sample reactivity measurements with pure fission products and structural material in five configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. The calculated neutron and adjoint spectra are in good agreement with former results obtained with the European route JEF2.2/ECCO/ERANOS. The C/E-values of the central reactivity worths of samples under investigation are given. Deviations in the results of both routes are due to the different libraries, codes, and self-shielding treatments used in the calculations. Results exceeding the experimental error are discussed. (author)

Some considerations for assurance of reactor safety from experiences in research reactors

International Nuclear Information System (INIS)

Okamoto, Sunao; Nishihara, Hideaki; Shibata, Toshikazu

1981-01-01

For the purpose of assuring reactor safety and strengthening research in the related fields, a multi-disciplinary group was formed among university researchers, including social scientists, with a special allocation of the Grant-in-Aid from the Ministry of Education, Science and Culture. An excerpt from the first year's report (1979 -- 1980) is edited here, which contains an interpretation of Murphy's reliability engineering law, a scope of reactor diagnostic studies to be pursued at universities, and safety measures already implemented or suggested to be implemented in university research reactors. (author)

Analysis and evaluation of ZPPR critical experiments for a 100 kilowatt-electric space reactor

International Nuclear Information System (INIS)

McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W.; Doncals, R.A.; Andre, S.V.; Porter, C.A.; Cowan, C.L.; Stewart, S.L.; Protsik, R.

1990-01-01

ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously needed for fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further reductions in the mass of the SP-100. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design

Multi-physic simulations of irradiation experiments in a technological irradiation reactor

International Nuclear Information System (INIS)

Bonaccorsi, Th.

2007-09-01

A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

Removal of uranium and salt from the Molten Salt Reactor Experiment

International Nuclear Information System (INIS)

Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

1998-01-01

In 1994, migration of 233 U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage

Removal of uranium and salt from the Molten Salt Reactor Experiment

Energy Technology Data Exchange (ETDEWEB)

Peretz, F.J.; Rushton, J.E.; Faulkner, R.L.; Walker, K.L.; Del Cul, G.D.

1998-06-01

In 1994, migration of {sup 233}U was discovered to have occurred at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). This paper describes the actions now underway to remove uranium from the off-gas piping and the charcoal bed, to remove and stabilize the salts, and to convert the uranium to a stable oxide for long-term storage.

Energy deposition measurements in fast reactor safety experiments with fission thermocouple detectors

International Nuclear Information System (INIS)

Wright, S.A.; Scott, H.L.

1979-01-01

The investigation of phenomena occurring in in-pile fast reactor safety experiments requires an accurate measurement of the time dependent energy depositions within the fissile material. At Sandia Laboratories thin-film fission thermocouples are being developed for this purpose. These detectors have high temperature capabilities (400 to 500 0 C), are sodium compatible, and have milli-second time response. A significant advantage of these detectors for use as energy deposition monitors is that they produce an output voltage which is directly dependent on the temperature of a small chip of fissile material within the detectors. However, heat losses within the detector make it necessary to correct the response of the detector to determine the energy deposition. A method of correcting the detector response which uses an inverse convolution procedure has been developed and successfully tested with experimental data obtained in the Sandia Pulse Reactor (SPR-II) and in the Annular Core Research Reactor

The Experience of Storage and Shipment for Reprocessing of HEU Nuclear Fuel Irradiated in the IRT-M Research Reactor and Pamir-630 Mobile Reactor

Energy Technology Data Exchange (ETDEWEB)

Sikorin, S. N.; Polazau, S. A.; Luneu, A. N.; Hrigarovich, T. K. [Joint Institute for Power and Nuclear Research–Sosny of the National Academy of Sciences of Belarus, Minsk (Belarus)

2014-08-15

At the end of 2010 under the Global Threat Reduction Initiative (GTRI), the Joint Institute for Power and Nuclear Research–“Sosny” (JIPNR–Sosny) of the National Academy of Sciences of the Republic of Belarus repatriated HEU spent nuclear fuel to the Russian Federation. The spent nuclear fuel was from the decommissioned Pamir-630D mobile reactor and IRT-M research reactor. The paper discusses the Pamir-630D spent nuclear fuel; experience and problems of spent nuclear fuel storage; and various aspects of the shipment including legal framework, preparation activities and shipment logistics. The conceptual project of a new research reactor for Belarus is also presented.

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

Science.gov (United States)

Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

2017-07-01

In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

Experience of developments and implementation of advanced fuel cycles of WWER-440 reactors

International Nuclear Information System (INIS)

Gagarinski, A.A.; Lizorkin, M.P.; Novikov, A.N.; Proselkov, V.N.; Saprykin, V.V.

2000-01-01

The paper presents the experience of development and implementation of advanced four- and five-year fuel cycles in the WWER-440 reactors, the results of experimental operation of the new fuel design and the main neutronic characteristics of the core. (Authors)

Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-01-01

The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

Criticality calculations in reactor accelerator coupling experiment (Race)

International Nuclear Information System (INIS)

Reda, M.A.; Spaulding, R.; Hunt, A.; Harmon, J.F.; Beller, D.E.

2005-01-01

A Reactor Accelerator Coupling Experiment (RACE) is to be performed at the Idaho State University Idaho Accelerator Center (IAC). The electron accelerator is used to generate neutrons by inducing Bremsstrahlung photon-neutron reactions in a Tungsten- Copper target. This accelerator/target system produces a source of ∼1012 n/s, which can initiate fission reactions in the subcritical system. This coupling experiment between a 40-MeV electron accelerator and a subcritical system will allow us to predict and measure coupling efficiency, reactivity, and multiplication. In this paper, the results of the criticality and multiplication calculations, which were carried out using the Monte Carlo radiation transport code MCNPX, for different coupling design options are presented. The fuel plate arrangements and the surrounding tank dimensions have been optimized. Criticality using graphite instead of water for reflector/moderator outside of the core region has been studied. The RACE configuration at the IAC will have a criticality (k-effective) of about 0,92 and a multiplication of about 10. (authors)

Supercritical Water Gasification of Biomass in a Ceramic Reactor: Long-Time Batch Experiments

Directory of Open Access Journals (Sweden)

Daniele Castello

2017-10-01

Full Text Available Supercritical water gasification (SCWG is an emerging technology for the valorization of (wet biomass into a valuable fuel gas composed of hydrogen and/or methane. The harsh temperature and pressure conditions involved in SCWG (T > 375 °C, p > 22 MPa are definitely a challenge for the manufacturing of the reactors. Metal surfaces are indeed subject to corrosion under hydrothermal conditions, and expensive special alloys are needed to overcome such drawbacks. A ceramic reactor could be a potential solution to this issue. Finding a suitable material is, however, complex because the catalytic effect of the material can influence the gas yield and composition. In this work, a research reactor featuring an internal alumina inlay was utilized to conduct long-time (16 h batch tests with real biomasses and model compounds. The same experiments were also conducted in batch reactors made of stainless steel and Inconel 625. The results show that the three devices have similar performance patterns in terms of gas production, although in the ceramic reactor higher yields of C2+ hydrocarbons were obtained. The SEM observation of the reacted alumina surface revealed a good resistance of such material to supercritical conditions, even though some intergranular corrosion was observed.

Characterization of the molten salt reactor experiment fuel and flush salts

International Nuclear Information System (INIS)

Williams, D.F.; Peretz, F.J.

1996-01-01

Wise decisions about the handling and disposition of spent fuel from the Molten Salt Reactor Experiment (MSRE) must be based upon an understanding of the physical, chemical, and radiological properties of the frozen fuel and flush salts. These open-quotes staticclose quotes properties can be inferred from the extensive documentation of process history maintained during reactor operation and the knowledge gained in laboratory development studies. Just as important as the description of the salt itself is an understanding of the dynamic processes which continue to transform the salt composition and govern its present and potential physicochemical behavior. A complete characterization must include a phenomenological characterization in addition to the typical summary of properties. This paper reports on the current state of characterization of the fuel and flush salts needed to support waste management decisions

Experience in Reviewing Small Modular Reactor Technology

International Nuclear Information System (INIS)

Ahmad Nabil Abdul Rahim; Alfred, S.L.; Phongsakorn, P.

2015-01-01

Malaysia is in the stage of conducting Preliminary Technical Feasibility Study for the Deployment of Small Modular Reactor (SMR). There are different types of SMR, some already under construction in Argentina (CAREM) and China (HTR-PM) - (light water reactor and high temperature reactor technologies), others with near-term deployment such as SMART in South Korea, ACP100 in China, mPower and NuScale in the US, and others with longer term deployment prospects (liquid-metal cooled reactor technologies). The study was mainly to get an overview of the technology available in the market. The SMR ranking in the study was done through listing out the most deployable technology in the market according to their types. As a new comer country, the proven technology with an excellent operation history will usually be the main consideration points. (author)

Oscillation experiments techniques in CEA Minerve experimental reactor

Energy Technology Data Exchange (ETDEWEB)

Antony, M.; Di-Salvo, J.; Pepino, A.; Bosq, J. C.; Bernard, D.; Leconte, P.; Hudelot, J. P.; Lyoussi, A. [CEA CADARACHE, DEN/DER/SPEx, 13108 Saint Paul-lez-Durance (France)

2009-07-01

This paper deals with experiments in the Minerve pool Zero Power Reactor. Minerve is mainly devoted to neutronics studies, in view to improve the calculation routes by reducing the uncertainties of the experimental databases for nuclides arising in plutonium and wastes management. Minerve experimental measurement programs are performed by using the oscillation technique. This experimental technique consists in a periodic insertion and extraction of samples containing the nuclide of interest in a well characterized neutron spectrum. The reactivity variation of the sample is compensated by a calibrated rotary automatic pilot using cadmium sectors. The normal accuracy for measurements of small-worth samples in Minerve by using such a technique is about 3% for absolute reactivity worth, including the uncertainties on the material balance and on the calibration step. Reactivity effects of less than 1.5 cent can be measured. The OSMOSE and the OCEAN programs have been carried out since 2005 and will last until 2011. These programs aim at improving, in different neutron spectra, the absorption cross sections of respectively a majority of the separated heavy nuclides from {sup 232}Th to {sup 245}Cm appearing during the reactor and the fuel cycle physics, and of current and future types of absorbers as Gd, Hf, Er, Dy and Eu. (authors)

PAHR experiments in the MELUSINE reactor

International Nuclear Information System (INIS)

Rousseau, D.; Dereymez, P.; Guyon, H.; Junod, E.; Ploujoux, M.; Tournebize, F.; Backs, H.

1983-01-01

After a hypothetical accident in a fast neutron reactor core, the nuclear fuel and construction materials melt partially. In several out-of-pile devices, the melting materials and the sodium coolant come to interact thermodynamically. In short, a few seconds after the accident a bed of debris immersed in sodium is formed on a plane of steel. The PAHR program has as principal objective to study the thermodynamic behaviour of this bed in the MELUSINE reactor, taking into account the most crucial parameters that rule the phenomena. More particularly, the aim is to draw attention to the bed behaviour beyond the fusion point of the steel up to the partial fusion of the fuel. The authors describe the CELIA capsule and its instrumentation; the operation conditions of the reactor and the coupling factor; the out-of-pile materials and their operation conditions. (Auth.)

Neutron optics experiments at the TRIGA Mark II reactor of the Atominstitut Wien

International Nuclear Information System (INIS)

Jericha, E.; Badurek, G.; Baron, M.; Hasegawa, Y.; Jaekel, M.; Klepp, J.; Rofner, A.; Sponar, S.; Trinker, M.; Villa, M.; Rauch, H.

2004-01-01

We present the layout and characteristics of the 3 neutron optics instruments located at the beam ports of the Vienna TRIGA reactor (hosted by the Atominstitut of the Austrian Universities, Vienna University of Technology) and the most recent experiments performed thereon. (author)

A WIMS E analysis of zero energy experiments performed on the Dragon reactor

Energy Technology Data Exchange (ETDEWEB)

Lancefield, M. J.; Broadhouse, B.; Woloch, F.

1974-10-15

UKAEA methods embodied in the WINS-E modular scheme of codes are described in their application to the analysis of zero energy experiments performed on the DRAGON reactor. Measured reactivity and reaction rate distributions are compared with the predictions of the analysis.

Slow control systems of the Reactor Experiment for Neutrino Oscillation

International Nuclear Information System (INIS)

Choi, J.H.; Jang, H.I.; Choi, W.Q.; Choi, Y.; Jang, J.S.; Jeon, E.J.; Joo, K.K.; Kim, B.R.; Kim, H.S.; Kim, J.Y.; Kim, S.B.; Kim, S.Y.; Kim, W.; Kim, Y.D.; Ko, Y.J.; Lee, J.K.; Lim, I.T.; Pac, M.Y.; Park, I.G.; Park, J.S.

2016-01-01

The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

Concept of the LORELEI Test Device for LOCA Experiment in the JHR Reactor

International Nuclear Information System (INIS)

Moran, N.; Ferry, L.; Azulay, A.; Mileguir, O.; Weiss, Y.; Szanto, M.

2014-01-01

Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis. Former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. In JHR material testing reactor, which is currently under construction, one significant experimental device is the LORELEI testing device. The objective is to examine the LOCA sequence influence on: thermo-mechanical behavior of the fuel clad, possible fuel relocation, corrosion at high temperature, oxidation, hydriding and resulted clad embrittment. The device is a single rod closed loop system placed on a displacement device inside a defined channel in the reflector. Several operational constrains on the device, as required by the reactor operational philosophy resulted quite a few challenges in the design. Constrains as: pre experimental re-irradiation phase under thermo-syphonic flow, application of active insulation to simulate the surrounding fuel, application of tensile force during refolding simulation, controlling the experiment with non-direct temperature measurement, etc. requires sophisticated solutions. The main objective of the conceptual design was to remove the uncertainties of those challenging requirements. The current presentation describes the approach applied defining the concept of the device, using sophisticated design combined with computational and experimental tools

Maintenance experience on reactor recirculation pumps at Tarapur Atomic Power Station

International Nuclear Information System (INIS)

Singh, A.K.

1995-01-01

Reactor recirculation pumps at Tarapur Atomic Power Station (TAPS) are vertical, single stage centrifugal pumps having mechanical shaft seals and are driven by vertical mounted 3.3 kV, 3 phase, 1500 h.p. electric motors. During these years of operation TAPS has gained enough experience and expertise on the maintenance of reactor recirculation pumps which are dealt in this article. Failure of mechanical shaft seals, damage on pump carbon bearings, motor winding insulation failures and motor shaft damage have been the main areas of concern on recirculation pump. A detailed procedure step by step with component sketches has helped in eliminating errors during shaft seal assembly and installation. Pressure breakdown devices in seal assembly were rebuilt. Additional coolant water injection for shaft seal cooling was provided. These measures have helped in extending the reactor recirculation pump seal life. Pump bearing problems were mainly due to failure of anti-rotation pins and dowel pins of bearing assembly. These pins were redesigned and strengthened. Motor stator winding insulation failures were detected. Stator winding replacement program has been taken up on regular basis to avoid winding insulation failure due to aging. 3 refs., 2 tabs., 7 figs

Investigation of neutrino properties in experiments at nuclear reactors: Present status and prospects

International Nuclear Information System (INIS)

Mikaelyan, L.A.

2002-01-01

The present status of experiments that are being performed at nuclear reactors in order to seek the neutrino masses, mixing, and magnetic moments, whose discovery would be a signal of the existence of physics beyond the Standard Model, is considered, along with their future prospects

Reactor Dynamics Experiments with a Sub-Critical Assembly

International Nuclear Information System (INIS)

Miley, G.H.; Yang, Y.; Wu, L.; Momota, H.

2004-01-01

A resurgence in use of nuclear power is now underway worldwide. However due to the shutdown of many university research reactors , student laboratories must rely more heavily on use of sub-critical assemblies. Here a driven sub-critical is described that uses a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The small IEC neutron source would be inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory

Heterogeneous computation tests of both substitution and reactivity worth experiments in the RB-3 reactor

International Nuclear Information System (INIS)

Broccoli, U.; Cambi, G.; Vanossi, A.; Zapellini, G.

1977-01-01

This report presents the results of several experiments carried out in the D 2 O-moderated RB-3 reactors at the CNEN's Laboratory of Montecuccolino, Bologna. The experiments referred to are either fuel-element substitution experiments or interstitial absorber experiments and were performed during the period 1972-1974. The results of measurements are compared with those obtained by means of computational procedure based on some ''cell'' codes coupled with heterogeneous codes. (authors)

Ageing management experience at NUR Reactor

Energy Technology Data Exchange (ETDEWEB)

Melllal, Sabrina; Rezig, Mohamed; Zamoun, Rachid; Ameur, Azeddin [Nuclear Research Center of Draria, Algiers (Algeria)

2013-07-01

NUR is a 1 MW, open pool reactor moderated and cooled by light water. It was commissioned in 1989. NUR is used for education and training in Nuclear Engineering and related topics for COMENA and National Scientific Community. It is also used to perform R and D works and services at national and regional levels. In this presentation, we describe the methodology and the main development activities related to the ageing management at NUR reactor. These activities include inspection actions and development actions to introduce modifications, to solve obsolescence issues in view to implement the required preventive and curative maintenance programs and to improve the performances of the installation. These actions involved mainly the Operation Assistance System of the Reactor (OAS), the secondary cooling loop, the cooling tower. A new OAS using a new technology and having more possibilities than the older one was introduced in the control system of the reactor. The OAS hardware structure, software structure and the main functions performed are presented. The second loop is entirely refurbished. Two new cooling towers are installed and connected to the main heat exchanger with new piping and valves. The architecture of this new installation is described and the performance assessed. Other actions which involve auxiliary systems like emergency electrical system, air pneumatic system and automatic fire extinguishing are presented.

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

2016-04-17

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

French experience and prospects in the reprocessing of fast breeder reactor fuels

International Nuclear Information System (INIS)

Megy, J.

1983-06-01

Experience acquired in France in the field of reprocessing spent fuels from fast breeder reactors is recalled. Emphasis is put on characteristics and quantities of spent fuels reprocessed in La Hague and Marcoule facilities. Then reprocessing developments with the realisation of the new pilot plant TOR at Marcoule, new equipments and study of industrial reprocessing units are reviewed [fr

UCN-VCN facility and experiments in Kyoto University Reactor

International Nuclear Information System (INIS)

Kawabata, Yuji; Okumura, Kiyoshi; Utsuro, Masahiko

1993-01-01

An ultracold and very cold neutron facility was installed in Kyoto University Reactor (KUR). The facility consists of a very cold neutron (VCN) guide tube, a VCN bender, a supermirror neutron turbine and experimental equipments with ultracold neutrons (UCN). The properties of each equipments are presented. UCN is generated by a supermirror neutron turbine combined with the cold neutron source operated with liquid deuterium, and the UCN output spectrum was measured by the time-of-flight method. A gravity analyzer for high resolution spectroscopy and a neutron bottle for decay experiments are now developing as the UCN research in KUR. (author)

Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

International Nuclear Information System (INIS)

Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

2014-01-01

The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

NRC data base for power reactor surveillance programs and for irradiation experiments results

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.

1991-01-01

The radiation damage of pressure vessel materials in nuclear reactors depends on many different factors, primarily fluence, fluence spectrum, fluence rate, irradiation temperature, and chemistry. These factors and, possibly, others such as heat treatment and type of flux used in weldments must be considered to reliably predict the pressure vessel embrittlement and to assure the safe operation of the reactor. Based on embrittlement predictions, decisions must be made concerning operating parameters, low-leakage fuel management, possible life extension, and the need for annealing of the pressure vessel. Large numbers of data obtained from surveillance capsules and test reactor experiments are needed, comprising many different materials and different irradiation conditions, to develop generally applicable damage prediction models that can be used for industry standards and regulatory guides. The US Nuclear Regulatory Agency has, therefore, sponsored a project to construct an Embrittlement Data Base (EDB) for a comprehensive collection of data concerning changes in material properties of pressure vessel steels due to neutron irradiation. A first version containing data from surveillance capsules of commercial power reactors, the Power Reactor Embrittlement Data Base (PR-EDB) Version 1, has been completed and is available to authorized users from the Radiation Shielding Information Center at the Oak Ridge National Laboratory. This document provides a discussion of the features of the current database. 1 fig

Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

International Nuclear Information System (INIS)

Althaus; Jansing; Kesseler; Kirchner; Menck

1974-01-01

Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

Report of blind start-up experiments carried out on the reactor Cabri between 4. and 8. July 1966

International Nuclear Information System (INIS)

Filipczak, N.; Filipczak, W.; Furet, J.; Kaiser, J.

1967-01-01

The blind start-up of a reactor without any neutronic data concerning a relatively wide range of power dynamics can be necessary when difficulties arise in the positioning of the detector or in neutron-gamma discrimination near the multiplying medium. The object of the experiments carried out on the reactor Cabri was to check the very complete analysis of the start-up accident which was studied on an analogue computer. The number of experiments carried out (12) is not sufficient to allow a definite conclusion. Nevertheless the blind start-up method advocated by N. FILIPCZAK and W. FILIPCZAK does not appear to be incompatible with the security during the operational phase (on condition that its dynamic characteristics are close to that of the reactor Cabri). (authors) [fr

Re-evaluation of the criticality experiments of the ''Otto Hahn Nuclear Ship'' reactor

Energy Technology Data Exchange (ETDEWEB)

Lengar, I.; Snoj, L.; Rogan, P.; Ravnik, M. [Jozef Stefan Institute, Ljubljana (Slovenia)

2008-11-15

Several series of experiments with a FDR reactor (advanced pressurized light water reactor) were performed in 1972 in the Geesthacht critical facility ANEX. The experiments were performed to test the core prior to its usage for the propulsion of the first German nuclear merchant ship ''Otto-Hahn''. In the present paper a calculational re-evaluation of the experiments is described with the use of the up-to date computer codes (Monte-Carlo code MCNP5) and nuclear data (ENDF/B-VI release 6). It is focused on the determination of uncertainties in the benchmark model of the experimental set-up, originating mainly from the limited set of information still available about the experiments. Effects of the identified uncertainties on the multiplication factor were studied. The sensitivity studies include parametric variation of material composition and geometry. The combined total uncertainty being found 0.0050 in k{sub eff}, the experiments are qualified as criticality safety benchmark experiments. (orig.)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

International Nuclear Information System (INIS)

Bacher, P.

1997-01-01

A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

Energy Technology Data Exchange (ETDEWEB)

LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

2014-07-01

Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

International Nuclear Information System (INIS)

Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

2010-01-01

plants dismantled to date in the US have repackaged the less activated waste back into the reactor vessel and shipped the entire assembly to the disposal site. Decisions like these can be driven by many factors such as disposal costs, transportation logistics, licensing fees, etc., but will have a significant impact on the segmentation and packaging plan so must be considered early in the planning phase. All segmentation tools are remotely controlled since the mechanical segmentation projects that Westinghouse has executed, so far, have been performed under water due to the high radiation levels. ALARA and personal safety is the number one priority during the site work. The complexity of the work requires well designed and reliable tools. Westinghouse has optimized the technologies from its experiences accumulated over the years. Its main focus has always been to improve tool handling and cutting speed, water cleanliness, fail-safe and safety aspects. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. The purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date. (authors)

French experience in operating pressurized water reactor power stations. Ten years' operation of the Ardennes power station

International Nuclear Information System (INIS)

Teste du Bailler, A.; Vedrinne, J.F.

1978-01-01

In the paper the experience gained over ten years' operation of the Ardennes (Chooz) nuclear power station is summarized from the point of view of monitoring and control equipment. The reactor was the first pressurized water reactor to be installed in France; it is operated jointly by France and Belgium. The equipment, which in many cases consists of prototypes, was developed for industrial use and with the experience that has now been gained it is possible to evaluate its qualities and defects, the constraints which it imposes and the action that has to be taken in the future. (author)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

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

Operating Experience Review(OER) and development of Issues Tracking System(ITS) for Jordan Research and Training Reactor(JRTR)

International Nuclear Information System (INIS)

Kim, Yong Jun; Lee, Hyun Chul

2011-01-01

The operation of the Jordan Research and Training Reactor which Korean consortium designs will start in March 2015. Though the power level of JRTR is different from the one of HANARO, a Korean research reactor, experience and expertise gained from the successful operation of the multipurpose research reactor, HANARO, would be applied for the design of JRTR because the basic operation principles of two reactors are almost same. From the point of human factors view, Operating Experience Review (OER) has the accurate purpose of reflecting accumulated knowledge to a new design and this activity are required to perform in the beginning stage of the control room designs in nuclear facilities. OER is to identify human factors engineering (HFE) issues related to safety. The issues from operating experience provide a basis for improving the plant design in a timely way. Identified issues are reported to an issues tracking system (ITS) so as to manage and resolve issues. HFE related safety issues are to be extracted from OER. The purpose of this paper is to present the scope and methods of OER for the JRTR design. In addition, a new ITS is proposed. The ITS is effective for issue management and has simplified states for issue development and small numbers of steps for issue control

Capacitor requirements for controlled thermonuclear experiments and reactors

International Nuclear Information System (INIS)

Boicourt, G.P.; Hoffman, P.S.

1975-01-01

Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested. (U.S.)

Capacitor requirements for controlled thermonuclear experiments and reactors

International Nuclear Information System (INIS)

Boicourt, G.P.; Hoffman, P.S.

1975-01-01

Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested

The fast reactor

International Nuclear Information System (INIS)

1980-02-01

The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

BNFL's experience in the sea transport of irradiated research reactor fuel to the USA

International Nuclear Information System (INIS)

Hudson, I.A.; Porter, I.

2000-01-01

BNFL provides worldwide transport for a wide range of nuclear materials. BNFL Transport manages an unique fleet of vessels, designed, built, and operated to the highest safety standards, including the highest rating within the INF Code recommended by the International Maritime Organisation. The company has some 20 years of experience of transporting irradiated research reactor fuel in support of the United States' programme for returning US obligated fuel from around the world. Between 1977 and 1988 BNFL performed 11 shipments of irradiated research reactor fuel from the Japan Atomic Energy Research Institute to the US. Since 1997, a further 3 shipments have been performed as part of an ongoing programme for Japanese research reactor operators. Where possible, shipments of fuel from European countries such as Sweden and Spain have been combined with those from Japan for delivery to the US. (author)

Reactor operator screening test experiences

International Nuclear Information System (INIS)

O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

1976-01-01

When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

Data sheets of fission product release experiments for light water reactor fuel, (2)

International Nuclear Information System (INIS)

Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo; Yamamoto, Katsumune; Nakazaki, Chozaburo.

1979-07-01

This is the second data sheets of fission products (FP) release experiments for light water reactor fuel. Results of five FP release experiments from the third to the seventh are presented: results of pre-examinations of UO 2 pellets, photographs of parts of fuel rod assemblies for irradiation and the assemblies, operational conditions of JMTR and OWL-1, variations of radioiodine-131 level in the main loop coolant during experimental periods, and representative results of post-irradiation examinations of respective fuel rods. (author)

Research reactor decommissioning experience - concrete removal and disposal -

International Nuclear Information System (INIS)

Manning, Mark R.; Gardner, Frederick W.

1990-01-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

Out-of-core detectors experiments in IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Abe, Alfredo Y.; Fuga, Rinaldo; Mendonca, Arlindo Gilson; Moreira, Joao M.L.; Angioletto, Elcio; Fanaro, Leda Cristina C.B.; Jerez, Rogerio; Coelho, Paulo R. Pinto; Santos, Adimir dos; Silva, Graciete S. de A. e; Diniz, Ricardo

2000-01-01

In order to study the response of out-of-core detectors, 16 stainless steel plates, with 0.5 cm thickness, were placed at the core-reflector interface of the IPEN/MB-01 reactor. BF 3 , 10 B and Au foil detectors were localized beyond the stainless steel plates in 7 different positions, one of them outside the moderator tank of the reactor for simulating a true PWR out-of-core detector. Calculations were performed for comparison with the experimental results with the TORT code, a three-dimensional transport theory discrete ordinate code. The experiment model utilized 16 energy groups, X-Y Z geometry, S 16 discrete ordinates and P 3 cross-sections. The obtained results showed a good agreement between measured and calculated reaction rates in Au foils. The larger discrepancy occurred for the case with 16 stainless steel with a 2,2% deviation. For position 7, outside of the moderator tank, the neutron flux was so low that it could not active the Au foils for the reaction rate measurements. (author)

Ultrasonic thermometry system for measuring very high temperatures in reactor safety experiments

International Nuclear Information System (INIS)

Carlson, G.A.; Sullivan, W.H.; Plein, H.G.; Kerley, T.M.

1979-06-01

Ultrasonic thermometry has many potential applications in reactor safety experiments, where extremely high temperatures and lack of visual access may preclude the use of conventional diagnostics. This report details ultrasonic thermometry requirements for one such experiment, the molten fuel pool experiment. Sensors, transducers, and signal processing electronics are described in detail. Axial heat transfer in the sensors is modelled and found acceptably small. Measurement errors, calculations of their effect, and ways to minimize them are given. A rotating sensor concept is discussed which holds promise of alleviating sticking problems at high temperature. Applications of ultrasonic thermometry to three in-core experiments are described. In them, five 10-mm-length sensor elements were used to measure axial temperatures in a UO 2 or UO 2 -steel system fission-heated to about 2860 0 C

Applying Bayesian neural networks to separate neutrino events from backgrounds in reactor neutrino experiments

International Nuclear Information System (INIS)

Xu, Y; Meng, Y X; Xu, W W

2008-01-01

A toy detector has been designed to simulate central detectors in reactor neutrino experiments in the paper. The samples of neutrino events and three major backgrounds from the Monte-Carlo simulation of the toy detector are generated in the signal region. The Bayesian Neural Networks (BNN) are applied to separate neutrino events from backgrounds in reactor neutrino experiments. As a result, the most neutrino events and uncorrelated background events in the signal region can be identified with BNN, and the part events each of the fast neutron and 8 He/ 9 Li backgrounds in the signal region can be identified with BNN. Then, the signal to noise ratio in the signal region is enhanced with BNN. The neutrino discrimination increases with the increase of the neutrino rate in the training sample. However, the background discriminations decrease with the decrease of the background rate in the training sample

New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Ma, X.B., E-mail: maxb@ncepu.edu.cn; Qiu, R.M.; Chen, Y.X.

2017-02-15

Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between {sup 235}U and {sup 239}Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller. - Highlights: • The covariance coefficients between isotopes vs reactor burnup may change its sign because of two opposite effects. • The relation between fission fraction uncertainty and atomic density are first studied. • A new MC-based method of evaluating the covariance coefficients between isotopes was proposed.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility

International Nuclear Information System (INIS)

Weiss, Frank-Peter; Dietze, Klaus; Jacqmin, Robert; Ishikawa, Makoto

2003-01-01

1 - Description: The RRR-SEG-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactor RRG-SEG (at RC Rossendorf / Germany) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in Note Technique SPRC/LEPh/93-230 (SEG) The SEG experiments are considered 'clean' integral experiments, simple and clear in geometry and well suited for calculation. In all SEG configurations only a few materials were used, most of these were standards. Due to the designed adjoint function (energy-independent or monotonously rising), the capture or scattering effect was dominant, convenient to check separately capture or scattering data. At first, analyses of the experiments have been performed in Rossendorf. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and

Experience and lessons learned in the assessment of safety justifications for experiments mounted in research reactors

International Nuclear Information System (INIS)

Cox, R.F.

1990-01-01

Some experiments in research reactors are arguably a risky undertaking due to their uncertain outcome. The justifications for such experiments require careful assessment to validate their undertaking. The public, the operators and the installation itself must be safeguarded. Assessment of the potential risk is an acquired skill but in doing so the route can be eased by learning from the lessons experience can teach. This paper, essentially for the usage of safety managers, sets out some of the issues relating to the assessment process gained from our experience over a few tens of years in the assessment of experiments. Many of the conclusions reached may appear all too obvious viewed in retrospect, but they were not necessarily clear at the time. Those organizations setting up assessment teams may find some of the conclusions of value such that their proposed management system can embrace methodologies for assessment that can avoid or lessen the impact of some of the pitfalls we have tried to identify. Failure to recognise some of these points may run the risk of delayed clearances, dilated timescales and cost overruns. It is in the hope of reducing all these penalties that we offer our experiences

Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Plevacova, K. [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Journeau, C., E-mail: christophe.journeau@cea.fr [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Piluso, P. [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Zhdanov, V.; Baklanov, V. [IAE, National Nuclear Centre, Material Structure Investigation Dept., Krasnoarmeiskaya, 10, Kurchatov City (Kazakhstan); Poirier, J. [CEMHTI, 1D, av. de la Recherche Scientifique, 45071 Orleans Cedex 2 (France)

2011-07-01

Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U{sub x}, Zr{sub y})O{sub 2-z} water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO{sub 2}, the zirconium carbide coating keeps its role of protective barrier with UO{sub 2}-Al{sub 2}O{sub 3} below 2000 deg. C but does not resist to a UO{sub 2}-Eu{sub 2}O{sub 3} mixture.

Experience in arranging shipments of spent fuel assemblies of commercial and research reactors

International Nuclear Information System (INIS)

Komarov, S.; Barinkov, O.; Eshcherkin, A.; Lozhnikov, V.; Smirnov, A.

2008-01-01

At present the key activities of Sosny Company are to inspect physical conditions, handle and arrange shipment of SFA including failed SFA. In 2003 after obtaining the license of Gosatomnadzor (Rostechnadzor now) entitled to handle nuclear materials in the process of their shipment, Sosny Company started preparing certification and arranging SFA shipment on its own. About 40 shipments of SFA were performed with participation of Sosny Company. Experience in handling failed SFA - an example of development of a new technology could be the transport and technological scheme of RBMK-1000 SFA shipment from Leningradskaya NPP that was designed by Sosny Company. TUK-11 cask was selected for this shipment. The example of change of transport and technological scheme is modification of the technology for handling and shipment of WWER-440 SFA from Kola NPP. Experience in arranging transportation - based on the results of development of logistics schemes for shipping SFA of reactor facilities Sosny Company justified and implemented composition of mixed trains containing rail cars of many types that enabled to perform shipment more efficiently in time and cost. Experience in arranging handling and shipment of research reactor SFA - over the past years the activity of Sosny Company was aimed at implementing international Russian Research Reactor Fuel Return (RRRFR) program. Since equipment of the majority of research centers doesn't allow for the large casks to be accepted and loaded, special casks of less mass and dimensions are used to ship SFA from research reactors. In RRRFR program it is assumed to use different casks for RR SFA such as Russian TUK- 19, TUK-128 and foreign SKODA VPVR/M and NAC-LWT. At present Sosny Company is involved in coordination of the efforts of the affected organizations in creating the type 'C' package for RR SFA in the RF. Conclusion: Under conditions of constant increase of the requirements to shipment safety and complication of regulations of all

SN transport analyses of critical reactor experiments for the SNTP program

International Nuclear Information System (INIS)

Mays, C.W.

1993-01-01

The capability of S N methodology to accurately predict the neutronics behavior of a compact, light water-moderated reactor is examined. This includes examining the effects of cross-section modeling and the choice of spatial and angular representation. The isothermal temperature coefficient in the range of 293 K to 355 K is analyzed, as well as the radial fission density profile across the central fuel element. Measured data from a series of critical experiments are used for these analyses

Experiences in the D ampersand D of the EBWR reactor complex at Argonne National Laboratory

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Boing, L.E.; Fellhauer, C.R.

1995-02-01

EBWR went critical in Dec 1957 at 20 MW(t), was upgraded to 100 MW(t) operation. EBWR was shut down July 1967 and placed in dry lay-up. In 1986, the D ampersand D work was planned in 4 phases: final planning and preparations for D ampersand D, removal of reactor systems, removal of reactor vessel complex, and final decontamination and project closeout. Despite precautions, there was an uptake of 241 Am by D ampersand D workers following underwater plasma arc cutting within the pool; the cause was traced to an experimental 241 Pu foil (200 μg) that was lost in the mid-1960s in the reactor vessel. Several major lessons were learned from this episode, among which is the fact that research facilities often involve unusual experiments which may not be recorded. Safety analysis and review procedure for D ampersand D operations need to be carefully considered since they represent considerably different situations than reactor operations. EBWR is one of the very few cases of a prototypic reactor facility designed, operated, tested and now D ampersand D'd by one organization

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Operation experience and maintenance at the TRIGA Mark II L.E.N.A. reactor

International Nuclear Information System (INIS)

Gngoli, F.; Berzero, A.; Lana, F.; Rosti, G.; Meloni, S.

2008-01-01

The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis, mostly for neutron activation analysis purposes. Moreover the reactor was completely shutdown in the first six months of this year to allow the dismantling of the NADIR experimental setup. The paper presents: - Reactor operation from July 1990 to June 1992; - Reactor users in the time period January 1990 - December 1991; - Specific activities of some radionuclides in the filling materials; - Specific activity of some radionuclides in thermal column materials. Operations related to dismantling of NADIR experimental facility are described. Finally the new thermal column configuration is presented. Starting from the end inside the reactor tank, a graphite layer (35 cm thick) was positioned, followed by a bismuth layer (10 cm thick) to reduce gamma-ray intensity. The old graphite rods were then positioned leaving in the central part, on the equatorial plane of the thermal column, a cavity whose vertical section has 40 cm width and 20 cm height. The bottom of the cavity, towards to the reactor tank, has been lined with additional layers of graphite (10 cm), bismuth (10 cm) and again graphite (1 cm). The new configuration allowed new experiments to be performed. The cavity in the central part has been created to allow the irradiation of large biological samples such as experimental animal and human livers. This is a peculiar step in a neutron capture boron therapy project to be carried out at the University of Pavia. In order to avoid an implemented 41 Ar production in the void space between shutters and the thermal column outer end, the external surface of the thermal column has been coated with boral sheets. The neutron flux profile, both thermal and epithermal, and cadmium ratio for gold are shown. The flux distribution appears to be adequate to proceed with the neutron capture boron therapy experiment. The LENA Health Physics Service has checked all phases of

Reactor physics

International Nuclear Information System (INIS)

Ait Abderrahim, H.

1998-01-01

Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

French experience in research reactor fuel transportation

International Nuclear Information System (INIS)

Raisonnier, Daniele

1996-01-01

Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packaging for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of spent fuel from power reactors are made on a regular and industrial basis, but this is not yet the case for the transport of spent fuel coming from research reactors. Each shipment is a permanent challenge and requires a reactive organization dealing with all the transportation issues. This presentation will explain the choices made by Transnucleaire and its associates to provide and optimize the corresponding services while remaining in full compliance with the applicable regulations and customer requirements. (author)

An evaluation of the results of the HTR fuel programme conducted in the Dragon reactor experiment

International Nuclear Information System (INIS)

Shepherd, L.R.

1982-01-01

The Dragon Reactor Experiment was used over a period of ten years to investigate the behaviour of HTR fuel elements under realistic service conditions. The purpose of the work was to develop fuel capable of meeting the requirements of commercial power reactors. The studies divided into areas concerned with the mechanical behaviour of the graphite core structure under fast neutron irradiation and the ability of the coated particle fuel to retain fissile products over commercially viable life-cycles. (author)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Management of historical waste from research reactors: the Dutch experience

International Nuclear Information System (INIS)

Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

2013-01-01

Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

International Nuclear Information System (INIS)

Grover, S.B.

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

The reactor antineutrino anomalies

Energy Technology Data Exchange (ETDEWEB)

Haser, Julia; Buck, Christian; Lindner, Manfred [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

2016-07-01

Major discoveries were made in the past few years in the field of neutrino flavour oscillation. Nuclear reactors produce a clean and intense flux of electron antineutrinos and are thus an essential neutrino source for the determination of oscillation parameters. Most currently the reactor antineutrino experiments Double Chooz, Daya Bay and RENO have accomplished to measure θ{sub 13}, the smallest of the three-flavour mixing angles. In the course of these experiments two anomalies emerged: (1) the reanalysis of the reactor predictions revealed a deficit in experimentally observed antineutrino flux, known as the ''reactor antineutrino anomaly''. (2) The high precision of the latest generation of neutrino experiments resolved a spectral shape distortion relative to the expected energy spectra. Both puzzles are yet to be solved and triggered new experimental as well as theoretical studies, with the search for light sterile neutrinos as most popular explanation for the flux anomaly. This talk outlines the two reactor antineutrino anomalies. Discussing possible explanations for their occurrence, recent and upcoming efforts to solve the reactor puzzles are highlighted.

In-pile experiments and test facilities proposed for fast reactor safety

International Nuclear Information System (INIS)

Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

1976-01-01

The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

Reactor Emergency Action Level Monitor: Knowledge acquisition experiences

International Nuclear Information System (INIS)

Touchton, R.A.

1987-01-01

This paper presents the Knowledge Acquisition experiences in developing the Reactor Emergency Action Level Monitor (REALM) Expert System Prototype. REALM is an expert system which interprets plant sensor data and provides advice on the proper emergency classification. The REALM project is being funded by the Electric Power Research Institute, Consolidated Edison is serving as the host utility, and the effort is being conducted by Technology Applications, Inc. REALM is being designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity, ultimately operating in a real-time, on-line processing environment. The paper discusses briefly the direct knowledge acquisition techniques used by the project team (who are themselves power industry engineers), to extract relevant knowledge from plant specifications and procedures

Sequential UASB and dual media packed-bed reactors for domestic wastewater treatment - experiment and simulation.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno

2016-01-01

A wastewater treatment system composed of an upflow anaerobic sludge blanket (UASB) reactor followed by a packed-bed reactor (PBR) filled with Sorbulite(®) and Polonite(®) filter material was tested in a laboratory bench-scale experiment. The system was operated for 50 weeks and achieved very efficient total phosphorus (P) removal (99%), 7-day biochemical oxygen demand removal (99%) and pathogenic bacteria reduction (99%). However, total nitrogen was only moderately reduced in the system (40%). A model focusing on simulation of organic material, solids and size of granules was then implemented and validated for the UASB reactor. Good agreement between the simulated and measured results demonstrated the capacity of the model to predict the behaviour of solids and chemical oxygen demand, which is critical for successful P removal and recovery in the PBR.

Safety concept of high-temperature reactors based on the experience with AVR and THTR

International Nuclear Information System (INIS)

Wachholz, Winfried; Kroeger, Wolfgang

1990-01-01

In the Federal Republic of Germany a reactor is considered safe if verification has been furnished that the requirements contained in paragraph 7 of the Federal German Atomic Energy Act are met for this reactor: demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose rate limits for normal operation and accidental conditions. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. In recent years, proposals for enhanced safety of nuclear power reactors or a radical change in safety philosophy have been made. This is characterised by 'inherently safe', 'super safe' and similar slogans. A quantitative definition of these requirements has not yet been established, but it is clear as a common objective that the event of beyond design basis accidents evacuation, relocation, and large scale contamination of ground should not occur. As a consequence of the Chernobyl accident the safety of all the NPPs in Germany has been reviewed. This analysis was completed for the THTR reactor in 1988. The same has been done for AVR reactor. The final evaluation of the HTR specific safety features have been fully confirmed. The HTR concepts under development are based on this experience. The HTR-Modul unit is currently being designed

Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

International Nuclear Information System (INIS)

Murata, Takashi; Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

2001-01-01

The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

Energy Technology Data Exchange (ETDEWEB)

Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

2001-12-01

The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

Observation of Energy and Baseline Dependent Reactor Antineutrino Disappearance in the RENO Experiment.

Science.gov (United States)

Choi, J H; Choi, W Q; Choi, Y; Jang, H I; Jang, J S; Jeon, E J; Joo, K K; Kim, B R; Kim, H S; Kim, J Y; Kim, S B; Kim, S Y; Kim, W; Kim, Y D; Ko, Y; Lee, D H; Lim, I T; Pac, M Y; Park, I G; Park, J S; Park, R G; Seo, H; Seo, S H; Seon, Y G; Shin, C D; Siyeon, K; Yang, J H; Yeo, I S; Yu, I

2016-05-27

The RENO experiment has analyzed about 500 live days of data to observe an energy dependent disappearance of reactor ν[over ¯]_{e} by comparing their prompt signal spectra measured in two identical near and far detectors. In the period between August of 2011 and January of 2013, the far (near) detector observed 31 541 (290 775) electron antineutrino candidate events with a background fraction of 4.9% (2.8%). The measured prompt spectra show an excess of reactor ν[over ¯]_{e} around 5 MeV relative to the prediction from a most commonly used model. A clear energy and baseline dependent disappearance of reactor ν[over ¯]_{e} is observed in the deficit of the observed number of ν[over ¯]_{e}. Based on the measured far-to-near ratio of prompt spectra, we obtain sin^{2}2θ_{13}=0.082±0.009(stat)±0.006(syst) and |Δm_{ee}^{2}|=[2.62_{-0.23}^{+0.21}(stat)_{-0.13}^{+0.12}(syst)]×10^{-3}  eV^{2}.

RA Reactor

International Nuclear Information System (INIS)

1989-01-01

This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

International feedback experience on the cutting of reactor internal components

International Nuclear Information System (INIS)

Boucau, J.

2014-01-01

Westinghouse capitalizes more than 30 years of experience in the cutting of internal components of reactor and their packaging in view of their storage. Westinghouse has developed and validated different methods for cutting: plasma torch cutting, high pressure abrasive water jet cutting, electric discharge cutting and mechanical cutting. A long feedback experience has enabled Westinghouse to list the pros and cons of each cutting technology. The plasma torch cutting is fast but rises dosimetry concerns linked to the control of the cuttings and the clarity of water. Abrasive water jet cutting requires the installation of costly safety devices and of an equipment for filtering water but this technology allows accurate cuttings in hard-to-reach zones. Mechanical cutting is the most favourable technology in terms of wastes generation and of the clarity of water but the cutting speed is low. (A.C.)

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

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

2016-11-01

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

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

Directory of Open Access Journals (Sweden)

Krása Antonín

2017-01-01

Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

Science.gov (United States)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

International Nuclear Information System (INIS)

Vierow, Karen

2008-01-01

This project is investigating countercurrent flow and 'flooding' phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the 'surge line' and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008

Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

International Nuclear Information System (INIS)

Primm, R.T. III; Mincey, J.F.

1982-01-01

The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations

Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

Directory of Open Access Journals (Sweden)

Yu-Feng Li

2014-11-01

Full Text Available We discuss reactor antineutrino oscillations with non-standard interactions (NSIs at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters.

Irradiation facilitates at the advanced test reactor

International Nuclear Information System (INIS)

Grover, Blaine S.

2006-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

Kinetic Study of COS with Tertiary Alkanolamine Solutions. 2. Modeling and Experiments in a Stirred Cell Reactor

NARCIS (Netherlands)

Littel, Rob J.; Versteeg, Geert F.; Swaaij, Wim P.M. van

1992-01-01

Absorption experiments of COS into aqueous solutions of MDEA and DEMEA at 303 K have been carried out in a stirred cell reactor. An absorption model, based on Higbie’s penetration theory, has been developed and applied to interpret the absorption experiments, using the kinetic data obtained in part

Operating Experience with the BR-5 Reactor; Experience acquise aupres du reacteur BR-5; Opyt ehkspluatatsii reaktora BR-5; Experiencia practica con el reactor BR-5

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A. I.; Kazachkovskij, O. D.; Pinkhasik, M. S.; Aristarkhov, N. N.; Karpov, A. V.; Larin, E. P.; Efimov, I. A.

1963-10-15

The paper discusses the carrying-out of repair and maintenance work on the radioactive liquid-metal circuit of the BR-5 fast neutron reactor. Attention is also given to problems of reactor operation after achievement of the planned 2% fuel burn-up with some disturbance of leak-tightness in individual fuel elements. An account is given of experience in discharging the active section, examining the condition and leak-tightness of the fuel elements, and decontaminating the equipment and piping of the first radioactive circuit after reaching 5% fuel burn-up. (author) [French] Dans ce memoire les auteurs decrivent l'execution des reparations et des travaux d'entretien dans le circuit radioactif liquide-metal du reacteur a neutrons rapides BR-5. Ils etudient egalement les problemes lies au fonctionnement du reacteur au taux de combustion de 2% prevu avec quelques defauts d'etancheite dans des elements combustibles particuliers. Ils decrivent le dechargementen zone active et examinent les conditions d'etancheite des elements combustibles. Ainsi que la decontamination de l'appareillage et des tuyauteries du premier circuit radioactif apres avoir atteint un taux de combustion de 5%. (author) [Spanish] En la memoria se examinan los problemas planteados por el mantenimiento del circuito radiactivo de metal liquido del reactor de neutrones rapidos BR-5. Se tratan cuestiones relacionadas con la explotacion del reactor una vez alcanzado el grado de combustion de 2%, previsto en el proyecto y luego de producirse ciertas alteraciones de la densidad de determinados elementos combustibles. Se describen la experiencia adquirida durante la descarga del cuerpo del reactor, las investigaciones del estado general y de la hermeticidad de los elementos combustibles y las operaciones de descontaminacion de la instalacion y de las tuberias del circuito radiactivo primario despues de alcanzado un grado de combustion de 5%. (author) [Russian] V doklade rassmatrivayutsya voprosy proizvodstva

Experience of joint reactor laboratory course with KUCA

International Nuclear Information System (INIS)

Nishina, Kojiro

1982-01-01

A description is given of a joint reactor laboratory course of graduate level, which is offered every summer since 1975 by nine associated japanese universities with the use of Kyoto University Critical Assembly (KUCA). A total of 315 students have taken the course in the last seven years. The course has been institutionalized with the background that it is extremely difficult for any single university in this country to have her own research or training reactor. By their effort the united faculty team of the course have succeeded in giving an effective, unique one-week course, taking advantage of their collaboration. By the scrutiny of student's responses to the course, one norices that a reactor is distinctively different from ordinary educational apparatus, in that the students do not play a main part in the adjustment of the apparatus itself. The beginners therefore tend to feel the reactor as a remote existence. This difficulty must be circumvented if an effective educational process is to be designed. (author)

Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

DEFF Research Database (Denmark)

Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, the mechanism is used to simulate published data on ignition delay time and laminar burning velocity of hydrogen. The flow reactor results show that at reducing, stoichiometric, and oxidizing conditions, conversion starts at temperatures of 750–775 K, 800–825 K, and 800–825 K, respectively. In oxygen atmosphere......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

Experiences in controlling the upgrading of TRIGA 2000 Bandung reactor

International Nuclear Information System (INIS)

Huda, K.; Wibowo, Y.W.; Suprawhardana, M.S.

2001-01-01

TRIGA 2000 Bandung Reactor was established in 1961 for research, education and isotope production purposes. The reactor reached its first criticality in October 1964 and operated at nominal power of 250 kW until 1971. In 1971 the reactor was upgraded to the power level of 1000 kW. In order to raise the capacity of isotope production, the reactor has been upgraded again to the power level of 2000 kW. During the modification of the reactor, the Center for Research and Development of Nuclear Techniques (CRDNT) was management of the reactor as it faced many problems, either technical or non-technical ones. This caused the upgrading activities to take a long time. At this time, the reactor upgrading has almost finished, and the nuclear commissioning is going on. Several aspects and problems associated with the upgrading process have been reviewed and the results are discussed in the present paper. (author)

Twelfth biennial conference on reactor operating experience maintenance and modifications for availability and efficiency

International Nuclear Information System (INIS)

1985-01-01

Separate abstracts were prepared for the papers presented in the following topical sessions: (1) Formal Availability Improvement Programs; (2) Maintenance Aids and Diagnostics; (3) Nuclear Plant Inspection and Testing; (4) Secondary Cycle Improvement to Programs; (5) Recent Plant Experiences; (6) Training, Research, and Test Reactors; (7) Impact of Plant Maintenance on Availability; (8) Intragranular Stress Corrosion Cracking: Repair, Inspection, and Mitigating Techniques; (9) Availability Improvement - Recent Development and Trends; (10) Water Chemistry and Corrosion Control; and (11) ALARA: Experiences with Dose Reduction

Project Experiences in Research Reactor Ageing Management, Modernization and Refurbishment. Report of a Technical Meeting on Research Reactor Ageing Management, Modernization and Refurbishment

International Nuclear Information System (INIS)

2014-08-01

Research reactors have played an important role in several scientific fields for around 60 years: in the development of nuclear science and technology; in the valuable generation of radioisotopes for various applications; and in the development of human resources and skills. Moreover, research reactors have been effectively utilized to support sustainable development in more than 60 countries worldwide. More than half of all operating research reactors are now over 40 years old, with many exceeding their originally conceived design life. The majority of operating research reactors face challenges due to the negative impacts of component and system ageing, which manifest in a number of forms. This situation was highlighted by a serious medical isotope supply crisis which peaked in mid-2010, when several major producing reactors underwent prolonged shutdowns due to extensive necessary overhauls of various systems. Several facilities have established a proactive systematic approach to managing ageing or mitigating its impact on safety and availability of isotopes. Others have tried to prevent or remedy the drawbacks of ageing on a case by case basis. Overall, a large body of knowledge related to ageing issues exists in many Member States. Collecting and sharing this information within the research reactor community can provide a solid foundation to develop a more systematic approach — that is, an ageing management programme to prevent negative consequences of ageing on the safety, and the operability and lifetime of operating, or even future, reactors. It may also help organizations to manage research reactors that have been in an extended shutdown state by ensuring that any required systems are operated and maintained in a safe manner prior to final decommissioning and disposal of fuel to safe storage facilities. Sharing experiences from projects undertaken to refurbish or replace equipment and systems, satisfy safety and regulatory requirements, improve

Project Experiences in Research Reactor Ageing Management, Modernization and Refurbishment. Report of a Technical Meeting on Research Reactor Ageing Management, Modernization and Refurbishment

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

Research reactors have played an important role in several scientific fields for around 60 years: in the development of nuclear science and technology; in the valuable generation of radioisotopes for various applications; and in the development of human resources and skills. Moreover, research reactors have been effectively utilized to support sustainable development in more than 60 countries worldwide. More than half of all operating research reactors are now over 40 years old, with many exceeding their originally conceived design life. The majority of operating research reactors face challenges due to the negative impacts of component and system ageing, which manifest in a number of forms. This situation was highlighted by a serious medical isotope supply crisis which peaked in mid-2010, when several major producing reactors underwent prolonged shutdowns due to extensive necessary overhauls of various systems. Several facilities have established a proactive systematic approach to managing ageing or mitigating its impact on safety and availability of isotopes. Others have tried to prevent or remedy the drawbacks of ageing on a case by case basis. Overall, a large body of knowledge related to ageing issues exists in many Member States. Collecting and sharing this information within the research reactor community can provide a solid foundation to develop a more systematic approach — that is, an ageing management programme to prevent negative consequences of ageing on the safety, and the operability and lifetime of operating, or even future, reactors. It may also help organizations to manage research reactors that have been in an extended shutdown state by ensuring that any required systems are operated and maintained in a safe manner prior to final decommissioning and disposal of fuel to safe storage facilities. Sharing experiences from projects undertaken to refurbish or replace equipment and systems, satisfy safety and regulatory requirements, improve

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

International Nuclear Information System (INIS)

Obara, Toru; Sekimoto, Hiroshi

1997-01-01

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

Reactor materials research as an effective instrument of nuclear reactor perfection

International Nuclear Information System (INIS)

Baryshnikov, M.

2006-01-01

The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

Operating experience of natural circulation core cooling in boiling water reactors

International Nuclear Information System (INIS)

Kullberg, C.; Jones, K.; Heath, C.

1993-01-01

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

Breeders: operational experience with fast power reactors in five states - more intensive German-French breeder cooperation

International Nuclear Information System (INIS)

Hueper, R.

1978-01-01

In the past year contracts have been signed and implemented for German-French cooperation in LMFBR development and commercialization. - The first German nuclear power station with a sodium cooled fast reactor, KNK II in Karlsruhe, is going into operation. - Construction of the prototype SNR 300 at Kalkar (Lower Rhine) is slowing down awaiting a decision of the German Federal Constitutional Court. - On the international level, remarkable experience in the operation of fast power reactors has accumulated. - Possible fuel cycle alternatives are being evaluated by an international committee. (orig.) [de

Leak-before-break experience in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Moan, G.D.; Coleman, C.E.

1988-01-01

In the Canada deuterium uranium (CANDU) reactor, each of the ∼ 400 hot pressure tubes containing the fuel bundles and the pressurized heat transport water is surrounded and insulated from the cold moderator by a calandria tube. The pressure tubes are made from cold-worked Zr-2.5 Nb with a minimum wall thickness of 4.19 mm, and the calandria tubes are made from annealed Zircaloy-2 with a minimum wall thickness of 1.37 mm. The annulus between these two tubes contains an inert gas. Leak-before-break has developed into an operational tool in CANDU reactors to prevent unstable failure of pressure tubes. A procedure for leak detection and reactor response has been developed from the use of the annulus gas, whose dew point is measured to ascertain if leaks have crept into the annulus. The characteristics of the crack are used to establish the response time for leak detection. The reactor is required to be shut down before the length of the slowly growing crack has reached the critical stage. This critical crack length, determined using slit burst tests on tubes, is the crack length at which the crack growth becomes unstable. The most likely crack growth mechanism is delayed hydride cracking. This mechanism requires three conditions to occur simultaneously: the material must be sensitive to delayed hydride cracking; zirconium hydrides must be present in the material; and the tensile stress must be sufficiently great

Description of the french graphite reactor and of the experiments performed in 1956; Presentation du premier reacteur a graphite francais et des experiences effectuees en 1956

Energy Technology Data Exchange (ETDEWEB)

Bussac, J; Leduc, C; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author) [French] Ce rapport presente les experiences qui furent faites sur le reacteur G1 et dont la description en detail fait l'objet des rapports suivants (670 'B a P'). Les principaux resultats sont fournis ici et commentes. On trouvera en outre les caracteristiques neutroniques du coeur actif de la pile, une description des principales installations et une mention des essais qui ont conduit au fonctionnement normal du reacteur en puissance. (auteur)

Electrical Capacitance Volume Tomography for the Packed Bed Reactor ISS Flight Experiment

Science.gov (United States)

Marashdeh, Qussai; Motil, Brian; Wang, Aining; Liang-Shih, Fan

2013-01-01

Fixed packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a highly desirable unit operation for long duration life support systems in space. NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. To validate these models, the instantaneous distribution of the gas and liquid phases must be measured.Electrical Capacitance Volume Tomography (ECVT) is a non-invasive imaging technology recently developed for multi-phase flow applications. It is based on distributing flexible capacitance plates on the peripheral of a flow column and collecting real-time measurements of inter-electrode capacitances. Capacitance measurements here are directly related to dielectric constant distribution, a physical property that is also related to material distribution in the imaging domain. Reconstruction algorithms are employed to map volume images of dielectric distribution in the imaging domain, which is in turn related to phase distribution. ECVT is suitable for imaging interacting materials of different dielectric constants, typical in multi-phase flow systems. ECVT is being used extensively for measuring flow variables in various gas-liquid and gas-solid flow systems. Recent application of ECVT include flows in risers and exit regions of circulating fluidized beds, gas-liquid and gas-solid bubble columns, trickle beds, and slurry bubble columns. ECVT is also used to validate flow models and CFD simulations. The technology is uniquely qualified for imaging phase concentrations in packed bed reactors for the ISS flight experiments as it exhibits favorable features of compact size, low profile sensors, high imaging speed, and

Prometheus Project Reactor Module Final Report, For Naval Reactors Information

International Nuclear Information System (INIS)

MJ Wollman; MJ Zika

2006-01-01

The Naval Reactors Prime Contractor Team (NRPCT) led the development of a power plant for a civilian nuclear electric propulsion (NEP) system concept as part of the Prometheus Project. This report provides a summary of the facts, technical insights, and programmatic perspectives gained from this two-year program. The Prometheus Project experience has been extensively documented to better position the US for future space reactor development. Major Technological and engineering challenges exist to develop a system that provides useful electric power from a nuclear fission heat source operating in deep space. General issues include meeting mission requirements in a system that has a mass low enough to launch from earth while assuring public safety and remaining safely shutdown during credible launch accidents. These challenges may be overcome in the future if there is a space mission with a compelling need for nuclear power to drive development. Past experience and notional mission requirements indicate that any useful space reactor system will be unlike past space reactors and existing terrestrial reactors

Gas-cooled reactors

International Nuclear Information System (INIS)

Schulten, R.; Trauger, D.B.

1976-01-01

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

Training courses at VR-1 reactor

International Nuclear Information System (INIS)

Sklenka, L.; Kropik, M.

2006-01-01

This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

Control-rod interference effects observed during reactor physics experiments with nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Itagaki, Masafumi; Miyoshi, Yoshinori; Gakuhari, Kazuhiko; Okada, Noboru; Sakai, Tomohiro.

1993-01-01

The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the G1 and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of G1 and G2 increased as a result of slight core burnup, about 1,400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects. (author)

The KASKA project - a Japanese medium-baseline reactor-neutrino oscillation experiment to measure the mixing angle $\\theta_{13}$ -

OpenAIRE

Kuze, Masahiro; Collaboration, for the KASKA

2005-01-01

A new reactor-neutrino oscillation experiment, KASKA, is proposed to measure the unknown neutrino-mixing angle $\\theta_{13}$ using the world's most powerful Kashiwazaki-Kariwa nuclear power station. It will measure a very small deficit of reactor-neutrino flux using three identical detectors, two placed just close to the sources and one at a distance of about 1.8km. Its conceptual design and physics reach are discussed.

Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

2008-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

Measurement of neutrinos released in nuclear reactors through the Borexino experiment; Mesure des neutrinos de reacteurs nucleaires dans l'experience Borexino

Energy Technology Data Exchange (ETDEWEB)

Dadoun, O

2003-06-01

The main goal of the Borexino experiment is to measure in real time the solar neutrino flux from the beryllium (Be{sup 7}) line at 862 keV. Beyond this pioneer low energy neutrino detection, Borexino will be able to measure solar neutrinos above the MeV, (B{sup 8} neutrinos and pep neutrinos), nuclear reactor neutrinos (with an average energy of 3 MeV) and the supernova neutrinos (their spectrum goes up to some ten MeV). In this work I mainly focus on the study of the nuclear reactors neutrinos. This field has recently been enriched by the results of the KamLAND experiment, which have greatly improved the determination of the neutrino oscillation parameters. In order to measure these events which are above the MeV, the Borexino collaboration entrusted the PCC group at College de France, with the tasks of developing a fast digit system running at 400 MHz: the FADC cards. The PCC group designed the FADC cards and completed them at the beginning of 2002. The first cards which were introduced in the main electronic acquisition unit allowed us to control their functioning and that of the acquisition software. FADC cards were also installed in the Borexino prototype, CTF. The data are analysed in order to determine a limit to the expected background noise of Borexino in measuring the nuclear reactor neutrinos. (author)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Operation experience with the TRIGA reactor Wien 2004

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-01-01

The TRIGA Mark-II reactor in Vienna is now in operation for more than 42 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent, the number of students participating in practical exercises has strongly increased, and also training courses for outside groups such as the IAEA or for the 2004 Eugene Wigner Course are using the reactor, because it is the only TRIGA reactor remaining in Austria. Therefore, there is no need for decommissioning and it is intended to operate it as long as possible into the next decade. Nevertheless, in early 2004 it was decided to prepare a report on a decommissioning procedure for a typical TRIGA Mark II reactor which lists the volumes, the activity and the weight of individual materials such as concrete, aluminium, stainless steel, graphite and others which will accumulate during this process (a summary of possible activated and contaminated materials and the activity of a single TRIGA fuel element as a function of fuel type and decay time in Bq is presented). The status of the reactor (instrumentation, fuel elements, cooling circuit, ventilation system, re-inspection and maintenance program, cost/benefit) is outlined. (nevyjel)

Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

International Nuclear Information System (INIS)

Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

2011-01-01

Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

Reactor safety

International Nuclear Information System (INIS)

Butz, H.P.; Heuser, F.W.; May, H.

1985-01-01

The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

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

Indian experience with radionuclide transport, deposition and decontamination in water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Narasimhan, S.V.; Das, P.C.; Lawrence, D.A.; Mathur, P.K.; Venkateswarlu, K.S.

1983-01-01

The present generation of water-cooled nuclear reactors uses construction materials chosen with utmost care so that minimum corrosion occurs during the life of the reactor. As interaction between the primary coolant and the construction materials is unavoidable, the coolant is chemically treated to achieve maximum compatibility. First measurements of the chemical and radiochemical composition of the crud present on the in-core and out-of-core primary heat transport system surfaces of a pressurized heavy-water-moderated and cooled reactor (PHWR) are given; then experience in India in the development of a low temperature, one-stage decontaminating formulation for chemical decontamination of the radioactive deposits formed on stainless steel surfaces under BWR conditions is discussed. The effect of the magnitude of the transients in parameters such as reactor power, system temperature, dissolved oxygen content in the coolant, etc. on the nature and migration behaviour of primary heat transport system crud in a PHWR is described. Contributions to radioactive sources and insoluble crud from different primary heat transport system materials are identified and correlated with reactor operations in a PHWR. Man-rem problems faced by nuclear reactors, especially during off-line maintenance, stress the need for reducing the deposited radioactive sources from system surfaces which would otherwise be accessible. Laboratory and on-site experimentation was carried out to effect chemical decontamination on the radioactive deposits formed on the stainless steel surfaces under BWR conditions. Both the reducing and oxidizing formulations were subsequently used in a small-scale, in-plant trial in the clean-up system of a BWR. More than 85% of the deposited 60 Co activity was found to have been removed by the oxidizing formulation. Efforts to develop a decontaminating mixture containing a reducing agent with the help of a circulating loop are in progress in the laboratory. (author)

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.)

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

Operating experiences of reactor shutdown system at MAPS

International Nuclear Information System (INIS)

Kotteeswaran, T.J.; Subramani, V.A.; Hariharan, K.

1997-01-01

The reactors in Madras Atomic Power Station (MAPS), Kalpakkam are Pressurised Heavy Water Reactors (PHWR) similar to RAPS, Kota. The moderator heavy water is pumped into the calandria from dump tank to make the reactor critical. Later with the calandria level held constant at 92% FT, the further power changes are being done with the movement of adjuster rods. The moderator is held in calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The shutdown of the reactor is effected by dumping the moderator water to dump tank by fast equalizing of helium gas pressure. In the revised mode of operation of moderator circuit after the moderator inlet manifold failure, the dump timing was observed to be more compared to the normal value. This was investigated and observed to be due to accumulation of D 2 O in the gas space above dump valves, which was affecting the helium equalizing flow. Also some of Indicating Alarm Meters (IAM) in protective system initiating the trip signals have failed in the unsafe mode. They have been modified to avoid the recurrence of the failures. (author)

Training of research reactor personnel

International Nuclear Information System (INIS)

Cherruau, F.

1980-01-01

Research reactor personnel operate the reactor and carry out the experiments. These two types of work entail different activities, and therefore different skills and competence, the number of relevant staff being basically a function of the size, complexity and versatility of the reactor. Training problems are often reactor-specific, but the present paper considers them from three different viewpoints: the training or retraining of new staff or of personnel already employed at an existing facility, and training of personnel responsible for the start-up and operation of a new reactor, according to whether local infrastructure and experience already exist or whether they have to be built up from scratch. On-the-spot experience seems to be an essential basis for sound training, but requires teaching abilities and aids often difficult to bring together, and the availability of instructors that does not always fit in smoothly with current operational and experimental tasks. (author)

Diagnostic measurement on research reactors

International Nuclear Information System (INIS)

Dach, K.; Zbytovsky, A.

A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

Calibration of RB reactor power

International Nuclear Information System (INIS)

Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

1976-09-01

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

Flooding Experiments and Modeling for Improved Reactor Safety

International Nuclear Information System (INIS)

Solmos, M.; Hogan, K.J.; VIerow, K.

2008-01-01

Countercurrent two-phase flow and 'flooding' phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing

Operational experience on sodium deposits in KNK reactor and RSB test facility

Energy Technology Data Exchange (ETDEWEB)

Jansing, W; Kirchner, G; Menck, J [INTERATOM, Bergisch Gladbach (Germany)

1977-01-01

A specific problem of sodium-cooled reactor plants is the formation of sodium aerosols which deposit at cold sections of the plant. Formation and behaviour of sodium aerosols depend on various factors. These may show extreme different effects under conditions which first seem to be identical. Thus, it is very difficult to set up general valid rules on sodium aerosols. By operational experience gained in different plants under divers operating conditions, knowledge is drawn which corresponds well with theoretical considerations. (author)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Determination of reactor parameters by single rod experiments

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Zdravkovic, Z; Ivkovic, M [Department of Reactor Physics and Dynamics, Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

1969-07-01

A method is developed for the experimental determination of reactor parameters by using an isolated fuel element. The method is based on the consideration of the fuel element as the source and sink of neutrons when placed in a constant neutron field. By measuring the perturbation of the original field produced by insertion of the test fuel element it was possible to determine the fuel element parameters defined by the heterogeneous reactor theory of Feinberg and Galanin as thermal neutron absorption constant {gamma}, and neutron multiplication constant {eta}. Statistical error for one series of measurement amount to 2% in the values of {eta} and {gamma}. The developed method was intended for the analysis of the nuclear characteristics of the fuel element in the stage of its construction and development for a given reactor system. (author)

Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

International Nuclear Information System (INIS)

Gieseler, Walter

1970-01-01

The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

Experiments on Critical Heat Flux for CAREM -25 Reactor

International Nuclear Information System (INIS)

Mazufri, C.M

2000-01-01

The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

The application of MVC design pattern in Daya bay reactor neutrino experiments online safety training system

International Nuclear Information System (INIS)

Liu Guanchuan; Chu Yuanping

2011-01-01

The article made an introduction to MVC, which is an architectural pattern used in software engineering. It specified the advantages and disadvantages of MVC and also the application of MVC in Daya Bay nuclear reactor neutrino experiment online safety training system. (authors)

Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J L; Aguilar H, F; Rivero G, T; Sainz M, E [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

2000-07-01

In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

Reactor cost driving items

International Nuclear Information System (INIS)

Spears, W.R.

1987-01-01

Assuming that the design solutions presently perceived for NET can be extrapolated for use in a power reactor, and using costing experience with present day fusion experiments and with fission power plants, the major components of the cost of a tokamak fusion power reactor are described. The analysis shows the emphasis worth placing on various areas of plant design to reduce costs

Applying Bayesian neural networks to event reconstruction in reactor neutrino experiments

International Nuclear Information System (INIS)

Xu Ye; Xu Weiwei; Meng Yixiong; Zhu Kaien; Xu Wei

2008-01-01

A toy detector has been designed to simulate central detectors in reactor neutrino experiments in the paper. The electron samples from the Monte-Carlo simulation of the toy detector have been reconstructed by the method of Bayesian neural networks (BNNs) and the standard algorithm, a maximum likelihood method (MLD), respectively. The result of the event reconstruction using BNN has been compared with the one using MLD. Compared to MLD, the uncertainties of the electron vertex are not improved, but the energy resolutions are significantly improved using BNN. And the improvement is more obvious for the high energy electrons than the low energy ones

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Over Twenty Years Of Experience In ITU TRIGA MARK-II Reactor

International Nuclear Information System (INIS)

Yavuz, Hasbi

2008-01-01

I.T.U. TRIGA MARK-II Training and Research Reactor, rated at 250 kW steady-state and 1200 MW pulsing power is the only research and training reactor owned and operated by a university in Turkey. Reactor has been operating since March 11, 1979; therefore the reactor has been operating successfully for more than twenty years. Over the twenty years of operation: - The tangential beam tube was equipped with a neutron radiography facility, which consists of a divergent collimator and exposure room; - A computerized data acquisition system was designed and installed such that all parameters of the reactor, which are observed from the console, could be monitored both in normal and pulse operations; - An electrical power calibration system was built for the thermal power calibration of the reactor; - Publications related with I.T.U. TRIGA MARK-II Training and Research Reactor are listed in Appendix; - Two majors undesired shutdown occurred; - The I.T.U. TRIGA MARK-II Training and Research Reactor is still in operation at the moment. (authors)

TREAT Reactor Control and Protection System

International Nuclear Information System (INIS)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

International Nuclear Information System (INIS)

Elsheikh, B.M.

2017-01-01

Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

Zero energy reactor 'RB'

Energy Technology Data Exchange (ETDEWEB)

Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1959-03-15

In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

On the Evaluation of Pebble Bead Reactor Critical Experiments Using the Pebbed Code

International Nuclear Information System (INIS)

Gougar, Hans D.; Sen, R. Sonat

2014-01-01

Critical experiments pose a particular but necessary challenge to validating pebble bed reactor design codes. Fuel and core heterogeneities, impurities in graphite, variable packing of pebbles, and moderately strong neutronic coupling are among the factors that inject uncertainty into the results obtained with lower fidelity core physics models. Some of these are addressed in this study. The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling

Experience with effluent release from the Omaha V. A. Hospital TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)