International Nuclear Information System (INIS)

... thermal power plants thermal reactors water cooled reactors WATER

International Nuclear Information System (INIS)

The invention concerns a procedure for operating reactors in nuclear power plants. It aims at utilizing power reserves in nuclear power plants. This can be achieved by a steam-side connection of the steam generators of two reactors. The amount of steam exchanged between the units is chosen in such a way that power changes at the steam turbines feedback mainly to the corresponding reactor. In order to realize a high power transfer it is necessary to return the amount of condensate produced in the steam receiving unit and corresponding to the power transferred to the feedwater system of the steam donating unit.

International Nuclear Information System (INIS)

Published in summary form only.

International Nuclear Information System (INIS)

This is the twenty-ninth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, and presents the most recent reactor data available to the IAEA. It contains the following summarized information: - General information as of the end of 2008 on power reactors operating or under construction, and shut down; - Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The IAEA's Power Reactor Information System (PRIS) is a comprehensive data source on nuclear power reactors in the world. It includes specification and performance history data of operating reactors as well as reactors under construction or reactors being ...

International Nuclear Information System (INIS)

... no. 2) p. 27-28. economics hydrogen power reactors nonmetals (ELEMENTS

International Nuclear Information System (INIS)

... excursions fftf reactor fuel elements lmfbr type reactors reactivity insertions

Science.gov (United States)

In this paper Particle Bed Reactor (PBR) designs are discussed which use /sup 233/U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of /sup 233/U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 ...

International Nuclear Information System (INIS)

The deliveration by the Nuclear Safety Commission was commenced on the alteration in reactor installation, as it had been inquired by the Ministry of International Trade and Industry. The alteration is the additional installation of the reactor No. 2 in the Sendai Nuclear Power Station, Kyushu Electric Power Co., Inc. It is a PWR power plant with thermal output of about 2,660 MW (electric output of 890 MW), to be installed, adjoining to the reactor No. 1 of the same type and capacity under construction. In the examination by MITI, it was confirmed that the technological capabilities for its construction and operation and the radiation protection measures in power generation are both sufficient. The contents of the examination include the siting conditions, the location and construction of reactor facilities, etc. ...

Energy Technology Data Exchange (ETDEWEB)

The operation of a nuclear power reactor necessarily implies the consumption or burnup of reactor fuel by fission and capture, which gives rise to a decrease in the reactivity of the reactor. The effect of americium formation on the criticality of a thermal power reactor using two types of fuel is studied. The three-dimensional core calculation is used to calculate the production of the transuranium isotopes and their effect on the effective multiplication factor (K[sub eff]). This effect cannot be neglected for thermal power reactors with UO[sub 2]-PuO[sub 2] fuel (3.11% after 70 weeks of operation). The effect of the transuranium isotopes on the K[sub eff] for a thermal power reactor with UO[sub 2] fuel is about 0.0018% and can be ignored. (author).

International Nuclear Information System (INIS)

To estimate the waste management needs of a fusion power reactor, a scheme for handling radioactive waste from a fusion plant has been devised. The handling scheme proceeds with radioactive waste, primarily from blanket replacement, being stored on-site; waste in cooled and shielded casks is then isolated off-site; finally, the materials are recycled. Using activities and component lifetimes supplied by designers, several conceptual fusion power reactors have been analyzed and their waste streams compared to fission reactors with regard to total activity, specific activity, and lifetimes of activity.

Energy Technology Data Exchange (ETDEWEB)

Comments are made on the Public Inquiry into CEGB's proposal to construct a pressurized water reactor (PWR) at Sizewell, UK. Aspects discussed include: time elapsed and its possible effect on the result; economics of nuclear power plants compared with coal-fired power plants; changes in real sterling/dollar exchange rates; effect of mineworkers' strike; the UK electric power generating system; AGR reactors compared with PWR reactors; extension of Magnox reactor life; radioactive waste management; political decisions.

Energy Technology Data Exchange (ETDEWEB)

Korea Atomic Energy Research Institute (KAERI) has been developing the integral reactor. The reactor head structure assembly (RHSA) is the structure installed over the reactor cover. Due to the characteristics of an integral reactor, there are many instrument cables and power cables coming out from the reactor cover and main components. The RHSA provides an interface location to connect these cables from Architecture Engineer (AE) and System Designer (SD). It also prevents a pipe whip and it prohibits instruments from becoming missiles. In this research, the axiomatic design approach for the RHSA is performed.

International Nuclear Information System (INIS)

A reference design was used in preparing a mathematical model of a fusion power plant with a tokamak reactor to investigate the extent to which the uncertainty still inherent in the physical reactor parameters affects the power costs. While only limited reductions of the power costs are achieved by improvements of the reference values for the reactor burn time, power density in the torus and load on the first wall, the power costs rise in keeping with the extent to which these parameters fall short of the reference values. As the results obtained in present-day experiments are still well below the reference values, a great deal of effort is still required in the fields of plasma physics and materials research to achieve an economically operating fusion power plant. (orig.).

International Nuclear Information System (INIS)

Nuclear physics parameters of the Wolsung CANDU-PHW reactor are computed by use of the PHWCELL computer code that is an improved version of LATREP. The PHWCELL code mainly computes cell parameters of heavy water moderated reactors, and modeling scheme of heavy water reactor cell calculations has been developed with the PHWCELL computer code. The reactor operating conditions considered in the study are cold zero power (CZP) and hot full power (HFP) with equilibrium poison. The cell parameters are also computed as a function of fuel burnup and the numerical results are compared with the results in PSR of the Wolsung unit and in the previous study. (author).

International Nuclear Information System (INIS)

SMART (System-Integrated Modular Advanced Reactor) is a first-of-the-kind integral reactor with 330 MW thermal power under active development by Korea Atomic Energy Research Institute (KAERI) for power generation and seawater desalination. SMART employs various design features that are not typically found in other nuclear power plants. Examples include a unique passive residual heat removal system (PRHRS), and enclosure of a pressurizer, eight helical steam generators, and eight canned reactor coolant pumps inside the reactor pressure vessel. This paper presents risk insights on the SMART reactor gained during the development of a regulatory PSA model by Korea Institute of Nuclear Safety (KINS)

Energy Technology Data Exchange (ETDEWEB)

The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year operational performance of ...

International Nuclear Information System (INIS)

A LOss of Electric Power(LOEP) experiment was conducted after a 30MW full power operation as one of the reactor performance tests to verify the design characteristics of the HANARO. The objective of LOEP test was to investigate the integral behaviors of the system and the components as well as the cooling characteristics when the electric power was lost unexpectedly. Through the test, it was confirmed that the residual heat from the core was safely removed by the natural convection cooling and the assistant power systems operated normally

Energy Technology Data Exchange (ETDEWEB)

Monte Carlo neutron transport methods can be used to verify the applicability of point kinetics for safety analysis of nuclear reactors. KENO-NR was used to obtain the transfer function of the Advanced Neutron Source reactor and the time delay between the core power production and the external detectors, a parameter of interest to the safety systems design. The good agreement between the Monte Carlo generated transfer function and the point kinetics transfer function validates that the uncommon ANS geometry does not preclude the use of point kinetics in the frequency range that was investigated. Various features of the power spectral densities also demonstrated the applicability of point kinetics. The time delay was obtained from the cross-power spectral density (CPSD) and is {approximately}15 ms. These analyses show that frequency analysis can be used experimentally to investigate ...

International Nuclear Information System (INIS)

2010 p. 11 Ukraine Inozemtsev, V. Department of Nuclear Energy, IAEA, Vienna

International Nuclear Information System (INIS)

Feb 2006 p. 80-84 Japan Mizuno, R. Matsuda, F. Japan Power Engineering

Energy Technology Data Exchange (ETDEWEB)

Canadian power and industrial projects world-wide are highlighted in this annual feature. A short section on the CANDU Wolsung Reactor in the Republic of Korea mentions that it went critical in November 1982 after taking only 60 months to complete.

International Nuclear Information System (INIS)

Wolsung nuclear power plant has experienced four occasions of reactor shutdown owing to heavy water leaks since its commercial operation. Among these heavy water leaks, only one case was acute and brought about reactor shutdown but the other cases listed below were chronic and repaired after manual reactor shutdown. (author). 4 tabs., 10 figs.

International Nuclear Information System (INIS)

Afterheat represents an important consideration in design of conceptual fusion power reactors, particularly during normal or unplanned shutdown. Afterheat calculations have been undertaken for various generic designs, but with special reference to the Culham DEMO reactor. These calculations have included the redistribution of heating by gamma ray transport. Selected temperature response calculations have been undertaken. (author).

International Nuclear Information System (INIS)

This paper explores the current trends in development of technology-neutral safety requirements to be used in the regulation of future nuclear power reactors and the role of the quantitative safety goals in the design of reactor safety systems. Establishing the requirements concerning the reliability of safety functions rather than on particular systems employed to achieve the functions, as well as the use of the recommendations of the International Commission on Radiological Protection (ICRP) on protection against potential exposure could form the basis of a technology-neutral framework for safety requirements on new reactor designs. Also it could contribute to international harmonisation of nuclear safety assessment practices as part of the licensing processes for future nuclear power plants. (author)

International Nuclear Information System (INIS)

A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author).

International Nuclear Information System (INIS)

A brief review of apparatuses used at enterprises engaged in industrial processing of spent nuclear fuel for dissolving dioxide nuclear fuel from power reactors is provided. Advantages and drawbacks of facilities operating in periodic, semi-continuous and continuous modes are considered. It is pointed out that today there are two promising trends in developments in the field, i.e. rotor- and vibrational-type dissolving apparatuses operated continuously

International Nuclear Information System (INIS)

A non-linear mathematical model of dynamics of horizontal steam generator for nuclear power unit with WWER type reactor is presented. To realize this model the GEMMA-120 simulation language for computer Odra-1204 has been used. Necessity of taking into account disposited thermal storage capacities along tubulation of a primary cycle is demonstrated. A number of lumped elements of reactor division against a required static accuracy of calculations has been determined. (author).

Science.gov (United States)

Utility interest has recently increased in potential future nuclear units that combine the characteristics of smaller size, greater simplicity, and more passive safety features. In response to such interest, General Electric (GE) began development in 1982 of a 600-MW(electric) reactor with simplified power generation and safety systems. This paper provides an overview of the simplified boiling water reactor (SBWR) design, with emphasis on the thermal-hydraulic aspects of the design. The SBWR is a natural circulation reactor requiring no pumps to circulate the water through the core.

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Energy Technology Data Exchange (ETDEWEB)

Encapsulation of a unique isotopic blend of krypton and xenon gas employs a special application of laser technology. The encapsulated gas is then used as the primary medium for detection and identification of failed nuclear fuel rods. The use of gas tagging as a means of detecting and identifying failed nuclear fuel rods has been successfully demonstrated and used by the Argonne National Laboratory, Experimental Breeder Reactor (EBR-2) Project, and the Westinghouse Hanford Company (WHC), Fast Flux Test Facility (FFTF) Fast Breeder Reactor Program. The Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan has selected this leak detection system for use in their MONJU Prototype Reactor fuel assemblies. The MONJU reactor is almost identical in design to the highly successful FFTF reactor, which is currently in standby status.

Energy Technology Data Exchange (ETDEWEB)

In CANDU reactors, the Regional Overpower Protection Trip (ROPT) system protects the reactor against overpowers in the reactor fuel, whether due to localized peaking within the core or a general increase in core power levels. Due to Primary Heat-Transport System (PHTS) aging the ROP trip setpoint is decreasing over time. Reductions in ROP trip setpoints are required to maintain the required trip-probability and ROP trip effectiveness, and results in a decrease of the ROP margin-to-trip during normal operation. In addition, full power operation can be threatened. In this point, to recover ROPT margin, channel power needs to be redistributed. ROPT setpoint is very conservative in normal operation because distortion of regional overpower is over 1.2 times as nominal power in slow loss of regulation (SLOR). Channel power ...

Energy Technology Data Exchange (ETDEWEB)

Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state ...

Energy Technology Data Exchange (ETDEWEB)

The Nuclear Safety Commission presented the report to the Minister of International Trade and Industry on April 5, 1984, after the careful investigation and deliberation on the alteration of installation of No.1 and No.2 reactor facilities in the Sendai Nuclear Power Station. The technical capability of Kyushu Electric Power Co., Inc., was recognized to be adequate. It was judged that the safety after this alteration of installation of the reactor facilities can be ensured. The main items of examination were as follows. The mechanical, nuclear and thermo-hydraulic designs of 17 x 17 B-type fuel assemblies were regarded as adequate. The coexistence of A-type and B-type fuel assemblies does not cause any problem about the safety. The safety at the time of abnormal transient change and accident in the mixed fuel assembly core was confirmed. In No.2 reactor, the degree of enrichment of ...

International Nuclear Information System (INIS)

Inertial Electrostatic Confinement (IEC) fusion was recently described by an Electric Power Research Institute (EPRI) review panel as potentially leading to a most attractive fusion reactor from a utility point of view, if the physics issues can be resolved. Consequently, a design for a small 25-MW electric D-"3He fueled power plant has been explored. Key power plant components consist of the IEC, direct energy conversion and a step-down converter for electrical power transmission. (author).

International Nuclear Information System (INIS)

The sequence of events from the assembly of a space nuclear power system to its integration in the Space Shuttle Transportation System (STS) is considered. First, the sequence followed for SNAP-10A, the only free world space reactor electric power system ever launched and operated in space, is reviewed. Before shipment, the SNAP-10A reactor was raised to operating temperature using electrically supplied heat and operated at low power for control calibration. Next we discuss shipment to the launch site, a phase that is critical because of the potential for various accidents. Once the power system arrives at the launch site, the processing sequence is performed. This sequence includes checkout, mating with the payload or upper stage launch vehicle, and integration into the STS.

Energy Technology Data Exchange (ETDEWEB)

The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the ...

International Nuclear Information System (INIS)

An Ottawa-based company, ECS-Power Systems Inc., has successfully completed a series of tests on an innovative device called a hydrodynamic port (HDP), which makes it possible to automatically initiate and maintain emergency cooling of a nuclear reactor core by natural processes, without relying in any way on human intervention, instrumentation, electric power, valves or moving parts of any kind.

International Nuclear Information System (INIS)

Weld metal, base material and stainless steel overlay specimens for Charpy tests and static tensile tests were irradiated for a year in a power reactor of the Bohunice nuclear power plant in place of the evaluated surveillance specimens. The material of the specimens was identical with that of the WWER-440 reactor pressure vessels, and was exposed to a fluence of (1.2 - 4.5) x 10"2"3 m"-"2 (E > 0.5 MeV) at approximately 270 degC. Some of the irradiated as well as unirradiated specimens were subjected to regeneration annealing at 475 degC for 168 h. The behavior of the materials after irradiation and annealing was evaluated. (author). 33 tabs., 32 figs., 8 refs.

Energy Technology Data Exchange (ETDEWEB)

To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously ...

Energy Technology Data Exchange (ETDEWEB)

This report discusses Test Campaign TC07 of the Kellogg Brown & Root, Inc. (KBR) Transport Reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). The Transport Reactor was operated as a pressurized gasifier during TC07. Prior to TC07, the Transport Reactor was modified to allow operations as an oxygen-blown gasifier. Test Run TC07 was started on December 11, 2001, and the sand circulation tests (TC07A) were completed on December 14, 2001. The coal-feed tests (TC07B-D) were started on January 17, 2002 and completed on April 5, 2002. Due to ...

Energy Technology Data Exchange (ETDEWEB)

The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant piping have led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National laboratory-sponsored research program aimed at investigating whether the probability of DEGB in Reactor Coolant Loop Piping of nuclear power plants is acceptably small such that the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimates the probability of indirect DEGB in Reactor Coolant piping as a consequence of seismic-induced structural failures within the containment of the GE supplied boiling water reactor at the ...

International Nuclear Information System (INIS)

Reactor Regulating System (RRS) of TAPP-3 and 4 (540 MWe PHWR) addresses issues of elaborate Flux Tilt Control as applied to large Reactor Cores in addition to the traditional Bulk Power (Actual Power) Control. The control of Bulk and Zonal Power by RRS through the use of Zonal Control Compartments (ZCCs) has been successfully demonstrated in the Indian PHWRs for the first time. Features like automation in Demand Power Maneuvering, Manual Movement of Reactivity Devices through the Human Machine Interface (HMI) and the supervised withdrawal of Shut-off Rods during Auto Criticality are also included. Special algorithms to measure and control the individual Zone Power and Bulk Power also form part of RRS algorithms. This paper describes the salient features of RRS of TAPP-3 and 4 and the improvement carried out based on the ...

International Nuclear Information System (INIS)

The activation aspects of pure fusion and hybrid fusion technology is studied to assess the radioactive safety of various fusion concepts including tokamak pure fusion, fissile fuel producing hybrid and radio waste transmuting hybrid. The activation properties of breeding, coolant and structural materials in fusion reactors might be quite different from those in fission reactors because of the high energy D-T fusion neutrons from the fusion cores. A study on the involved activation reactions and the uncertainties of the associated nuclear cross-sections is carried. The activation properties of various first wall concepts and blanket concepts are discussed. The radioactive inventory during the operation lifetime and the potential hazard of the radioactive nuclides with respect to near term (reprocessing) and long term (waste disposal) aspects are calculated, with reference to ITER/NET (International Thermonuclear Experiment ...

International Nuclear Information System (INIS)

Papers are presented on space power requirements and issues, space photovoltaic systems, space solar dynamic systems, space thermal systems, manned and unmanned space power systems, thermionics, and thermoelectrics. Also considered are high power devices for space power systems, high power conversion for space power systems, 1-10 kWe nuclear space power sources, 100-kW class nuclear power concepts, space reactor safety, and multimegawatt space nuclear power systems. Other topics include space power systems automation, space kilovolt technology, space power electronics, space lithium and nickel-cadmium batteries, lithium sodium storage, and space fuel cells. Papers are also presented on space nickel hydrogen batteries, alternative energy ...

International Nuclear Information System (INIS)

These proceedings represent papers presented at the 12th symposium on Space Nuclear Power and Propulsion held in Albuquerque, New Mexico. The symposium theme was ''commercialization and technology transfer''. The topics discussed include: wireless power transmission, solar power from space next generation spacecraft, space power electronics and power management, flight testing of components, manufacturing and processing of materials, nuclear propulsion, reactors and shielding and many others of interest to the scientific community representing industry, government and academic institutions. There were 163 papers presented at the conference and 60 have been abstracted for the Energy Science and Technology database.

Science.gov (United States)

This report presents a general investigation of the transport requirements associated with the construction and operation of conceptual fusion reactors. Projections of amounts of construction and operating materials requiring transportation are presented for several proposed designs. The material to be shipped is described along with the shipping containers that might be used, the transport modes and the expected impact of transporting these materials. Transportation of both radioactive and nonradioactive materials will be required. Most of these materials are routinely shipped by the transportation industry. Transportation requirements of a representative fusion reactor are also compared with Liquid Metal Fast Breeder Reactor (LMFBR) requirements.

Energy Technology Data Exchange (ETDEWEB)

Late last year the Wolsung Candu in Korea suffered an incident which resulted in heavy water being released from the primary system into the containment. With the unit now back at full power, this article examines the causes of the incident and the action which is being taken to prevent it happening again.

Science.gov (United States)

Liquid Phase Catalytic Exchange (LPCE)- Cryogenic Distillation(CD) process was studied which could be available for an optimal tritium removal process of pressurized heavy water reactor system at Wolsung nuclear power plant in the near future. Based upon ...

International Nuclear Information System (INIS)

The Daya Bay reactor neutrino experiment is designed to study the disappearance of antineutrinos from the Daya Bay nuclear power plant in China. The goal of this experiment is to measure the remaining unknown neutrino mixing parameter ?13 with high precision: sin2(2?13)<0.01. The experiment is presently under construction and it is anticipated that data acquisition will begin in 2011.

International Nuclear Information System (INIS)

The temperature coefficient has been investigated on the Wolsung nuclear power reactor, in which fuel is natural uranium dioxide and moderator heavy water. The numerical computations are carried out in terms of changes of the effective neutron multiplication factor with respect to fuel, moderator, and coolant temperatures. Those results are compared with the computed values of temperature coefficient based on the LATREP computer code. (author).

Energy Technology Data Exchange (ETDEWEB)

Low power wire activations are being performed in the Oak Ridge Research Reactor (ORR) as part of the whole-core LEU demonstration experiments. Calculations of the demonstration cores, including simulation of the wire activations, are being performed at Argonne National Laboratory (ANL). This paper presents the results of comparisons for 293 wires from five cores and shows that, on the average, the integrated activities agree within 6%.

Energy Technology Data Exchange (ETDEWEB)

This report discusses a desk-top computer program has been developed for estimating the costs, waste volumes, and occupational radiation exposures associated with decommissioning light-water reactor power stations. Cost categories and cost algorithms used in the program are discussed and a brief description of the user interface is given.

Science.gov (United States)

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the ...

International Nuclear Information System (INIS)

The Daya Bay reactor neutrino experiment is located at the Daya Bay nuclear power plant in Shenzhen, China. The experiment deploys eight 'identical' antineutrino detectors to measure antineutrino fluxes from six 2.9 GWth reactor cores in three underground experimental halls at different distances. The target zone of the Daya Bay detector is filled with 20 t 0.1% Gd doped LAB liquid scintillator. The baseline uncorrelated detector uncertainty is ?0.38% using current experimental techniques. Daya Bay can reach a sensitivity of sin22?13<0.01 with baseline uncertainties after 3 years of data taking.

International Nuclear Information System (INIS)

The potential public safety impacts from accidents in conceptual fusion power plants were investigated. Fusion was found to have some potential for accidents, as does any energy generating system. Functions of fusion power plants were identified that possess sufficient potential for an accidental release of toxic materials to the environment. An assessment was made of the impact of the potential accidents and recommendations are included for R and D that will allow incorporation of safety concerns in fusion power plant design. This work was based on a review of information available in conceptual design documents of fusion reactor systems.

International Nuclear Information System (INIS)

SMART (System integrated Modular Advanced ReacTor) , is under development at the Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection system (SIS), and an adoption of 4 trains of passive residual heat removal system (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a steam generator tube rupture (SGTR) is one of the most important initiating events which results in a high core damage frequency. Clear understanding of accident ...

Energy Technology Data Exchange (ETDEWEB)

A general space- and energy-dependent formalism is developed in order to analyze zero-power neutron noise experiments in fast reactor systems. A generalized dispersion equation is combined with theoretical expressions for the experimentally measured power spectral density and variance-to-mean ratio which makes it possible to express these quantities in terms of a double moment of the Laplace and Fourier transformed Green's function of a slowing-down operator rather than those of the full Boltzmann operator. Several spatial approximations are analyzed in the context of the general formalism. In each case, the power spectral density and variance-to-mean ratio are written in terms of an appropriate fast reactor dispersion law for the medium which can be calculated from the solution to a simple slowing-down equation. The resultant expression for the power ...

International Nuclear Information System (INIS)

In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

International Nuclear Information System (INIS)

This paper describes an electronic imaging system for producing real time neutron radiography from a low power research reactor, which will allow inspections of samples with high efficiency, in terms of measuring time and result analysis. This system has been implanted because of its potential use in various scientific and industrial areas where neutron radiography with photographic film could not be applied. This real time system is installed in neutron radiography facility of Argonauta nuclear research reactor, at the Instituto de Engenharia Nuclear of the Comissao Nacional de Energia Nuclear, in Brazil. It is adequate to perform real time neutron radiography of static and dynamic events of samples.

International Nuclear Information System (INIS)

Tarapur Atomic Power Station unit-4 is first 540 MWe pressurized heavy water reactor. Heavy water is used as the coolant and moderator. With reactor operation tritium is formed by absorption of neutron by Deuterium atoms. Experience in the 220 MWe indicates that tritium is main contributor to the internal dose of radiation workers. Study on Tritium build up in Primary Heat Transport (PHT) and Moderator (MOD) system was carried out at the initial stage of the operation of the unit-3 and 4. This paper brings out tritium activity buildup in the PHT and MOD systems and its comparison with 220 MWe reactors. This paper helps in estimation of the internal dose contribution to the radiation workers at TAPS 3 and 4. (author)

British Library Electronic Table of Contents (United Kingdom)

An experimental investigation on the thermal decomposition of CH4 into C and H2 was carried out using a 5kW particle-flow solar chemical reactor tested in a solar furnace in the 1300-1600K range. The reactor features a continuous flow of CH4 laden with mm-sized carbon black particles, confined to a cavity receiver and directly exposed to concentrated solar irradiation of up to 1720 suns. The reactor performance was examined for varying operational parameters, namely the solar power input, seed particle volume fraction, gas volume flow rate, and CH4 molar concentration. Methane conversion and hydrogen yield exceeding 95% were obtained at residence times of less than 2.0s. A solar-to-chemical energy conversion efficiency of 16% was experimentally reached, and a maximum value of 31% was numer...

Energy Technology Data Exchange (ETDEWEB)

The ORR operated at an average power level of 29.7 MW for 85.3% of the time during this period. The reactor was shut down on fifteen occasions, nine of which were unscheduled. Reactor downtime needed for refueling and checks was normal. The reactor remained available for operation 88.3% of the time. Special tests completed during the quarter included: (1) transfer of LEU fuel elements CLE-202 and NLE-201 from core positions B-9 and B-2 to core positions C-5 and C-6 for continued operation; and (2) calculation of maximum heat flux in LEU elements CLE-201 and NLE-202 in core positions A-2 and A-8. In-service inspections included inspections of ORR decay tank, primary heat exchanger No. 4, and the 24-in. strainer.

Energy Technology Data Exchange (ETDEWEB)

The extent to which the size of a modular stellarator reactor may be reduced is investigated by means of an analytic model of the reactor. The various means employed include varying the blanket/shield thickness, the power output and the wall loading. An optimum design is found, the major radius of which tends to be insensitive to changes in these quantities, although a decrease in the power output leads to a rather smaller decrease in reactor dimensions, as would be expected. Varying the plasma beta at fixed (iota/2..pi..)/sup 2/epsilon or, alternatively, increasing the rotational transform per field period, may, however, allow configurations with fewer field periods to be accessed which have a substantially smaller major radius than the 'standard case' adopted. The magnetics of various configurations required by the model are checked by field line following and ...

International Nuclear Information System (INIS)

Several configurations of moderating and shielding materials have been designed and measured on the LVR-15 reactor for boron neutron capture therapy (BNCT) purposes. To determine the neutron and gamma ray space-energy distributions in the cylindrical geometry, the two-dimensional code DOT with the coupled neutron-gamma data library DLC-36 was used. The experimental verification of the beam parameters was performed in the LVR-15 reactor thermal column empty space with layers of graphite, aluminium, alumina, lead and bismuth. Attention was paid to establishing techniques and instrumentation for monitoring the neutron and gamma ray dose and beam quality. The thermal and epithermal flux densities were measured by activation foils, the neutron spectrum was determined with a Bonner spectrometer and gamma ray background with a scintillation spectrometer. The distribution of thermal neutrons in the human head phantom was mapped with a small ...

International Nuclear Information System (INIS)

The main problems arising in decommissioning nuclear-powered submarines (NPS) relate to choosing a concept of handling reactor compartments followed by handling technology development. Reactor compartments (RC) are characterized with extremely space-saving or integral layout of large-size power equipment and systems, restricted access for dismantling, high radiation dose rates in a number of bays of RC. The above RC features pose a problem to find optimum option of RC utilization which on the one hand would be the most cost efficient, and the safest as possible on the other, i.e. dose commitments of personnel involved should be minimum, and effect on population and environment should be negligible. The main radiation factors specifying safety in RC handling at any decommissioning stage are as follows: (1) total radioactivity integrated in reactor facility (RF); (2) distribution of ...

International Nuclear Information System (INIS)

Full text: Research reactors have been playing a multi dimensional role in areas of nuclear fuel cycle programme, radio-isotope productions, neutron beam research etc. To ensure an efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are required on routine basis. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation requires a prior estimation of the reactivity load due to the sample, heating rate and the activity developed in it during irradiation. For the safety of the personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be ...

International Nuclear Information System (INIS)

Apppendix B of 10 CFR Part 50 establishes quality assurance requirements for the operation of nuclear power plant safety-related structures, systems and components. This Guide describes an acceptable method for complying with these regulations with regard to overall quality assurance program requirements for the operation phase of nuclear power plants. Input to this Guide has been provided by the Advisory Committee on Reactor Safeguards.

Energy Technology Data Exchange (ETDEWEB)

The evaluation of the hazard posed to the skin by very small radioactive sources (diameter < 1mm) has become popularly known as the 'hot particle' problem in European and American nuclear reactor facilities. In this study, research to detect hot particle was performed in Wolsung Nuclear power plant (NPP) in Korea.

Energy Technology Data Exchange (ETDEWEB)

The special geodetic problems encountered during the construction of the Paks nuclear power plants are treated. The main building with its hermetically connected components including the reactor, the steam generators, the circulation pumps etc. impose special requirements on the control net of datum points. The geodesy tasks solved during the construction of the main building are presented in details.

Energy Technology Data Exchange (ETDEWEB)

This standard addresses the design bases for light water reactor, nuclear power plant structures and components essential for the protection of public health and safety from the potential adverse effects of pipe whip, jet impingement, pressurization of compartments outside containment, environmental conditions and flooding associated with a postulated pipe rupture. The design bases for missile protection and the design bases for containment pressurization are not within this standard.

Energy Technology Data Exchange (ETDEWEB)

Nuclear power has been considered for space applications since the 1960s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors/ rocket-engines in the Rover/NERVA programs. However, changes in environmental laws may make the redevelopment of the nuclear rocket more difficult. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel form significantly different from NERVA may be needed to ensure public support. The Center for Space Nuclear Research (CSNR) is pursuing development of tungsten based fuels for use in a NTR, for a surface power reactor, and to encapsulate radioisotope power sources. The CSNR Summer Fellows program has investigated the feasibility of several missions enabled by the NTR. The potential mission benefits of a nuclear rocket, historical achievements of the previous programs, and recent ...

Science.gov (United States)

In March 2007 in-pile testing of LEU U-Mo full-size IRT type fuel elements was started in the MIR reactor. Four prototype fuel elements for Uzbekistan WWR SM reactor are being tested simultaneously - two of tube type design and two of pin type design. The dismountable irradiation devices were constructed for intermediate reloading and inspection of fuel elements during reactor testing. The objective of the test is to obtain the experimental results for determination of more reliable design and licensing fuel elements for conversion of the WWR SM reactor. The heat power of fuel elements is measured on-line by thermal balance method. The distribution of fission density and burn-up of uranium in the volume of elements are calculated by using the MIR reactor MCU code (Monte-Carlo) model. In this paper the design of fuel elements, the technique, main parameters and ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) was designed to test fast-reactor fuels and other nonfuel materials. In its 37 reactor cycles of operations, the FFTF reactor has performed very well and successfully completed all the irradiation testings with an operating efficiency factor as high as 98%. Since FFTF is an experimental reactor, its core loading changed from cycle to cycle. Depending on the number of test assemblies in the core and their location, the core loading can change significantly from an essentially homogeneous core loading to a relatively nonhomogeneous or even highly localized heterogeneous loading. Consequently, the core reload design and initial criticality analyses were required for each operating cycle. The zero power initial critical control rod bank height was predicted before each reactor startup. The initial critical prediction depends on the ...

International Nuclear Information System (INIS)

The pressurized Heavy Water Reactor (PHWR) consists of horizontal calandria vessel containing a large number of pressure tubes (fuel channels) connected to the reactor inlet and outlet headers by individual feeders. Coolant flow distribution among the pressure tubes play a vital role in extraction of thermal power. For these reactors one of the design objectives is to achieve uniform coolant outlet temperatures by providing coolant flows according to the channel power. This is achieved by the design process known as feeder sizing. This basically consists of accounting for the individual channel power and centre line geometry of individual feeder and iteratively adjusting the feeder hydraulic resistances within the design constraints such as limiting flow velocities, channel flows. Recently, the first unit of 540 MWe i.e Tarapur Atomic Power ...

International Nuclear Information System (INIS)

The second regular inspection of No.2 plant in Sendai Nuclear Power Station was carried out from October 5, 1987 to January 8, 1988. the parallel operation was resumed on December 8, 1987, 65 days after the parallel off. The facilities as the object of inspection were the reactor proper, reactor cooling system, measurement and control system, fuel facilities, radiation control facilities, waste facilities, reactor containment installation, and emergency power generation system. On these facilities as the object of inspection, the appearance, disassembling, leak, function, performance and other inspections were carried out, and abnormality was not found at all. The works related to this regular inspection were accomplished within the range of allowable dose based on the relevant laws. The main reconstruction works carried out during the period of this regular inspection were the ...

International Nuclear Information System (INIS)

At Tarapur Atomic Power Station 3 and 4, 540 MWe Pressurised Heavy Water Reactors, core being large in size requires a continuous in core monitoring for local flux disturbances. Nearly 200 Self Powered Neutron Detectors (SPNDs) of the Straight Individually Replaceable (SIR) type are distributed in the reactor core. For purpose of reactor regulation and protection, cobalt SPNDs that have a prompt response for changes in power is used for in-core flux mapping, vanadium SPNDs that provide accurate measure of neutron flux, even though having slow response is used In core SPNDs are placed in Vertical Flux Units (VFU) and Horizontal Flux Units (HFUs). These SPNDs were to be replaced at regular intervals to meet the design intent. Cobalt SPNDs have dose rates of the order of 1000 Gy/h and the Mineral Insulated (MI) cables of Vanadium SPNDs have dose rates of the order ...

Energy Technology Data Exchange (ETDEWEB)

The economic prospects for magnetic fusion energy can be dramatically improved if for the same total power output the fusion neutron first-wall (FW) loading and the system power density can be increased by factors of 3 to 5 and 10 to 30, respectively. A number of compact fusion reactor embodiments have been proposed, all of which would operate with increased FW loadings, would use thin (0.5 to 0.6 m) blankets, and would confine quasi-steady-state plasma with resistive, water-cooled copper or aluminum coils. Increased system power density (5 to 15 MWt/m/sup 3/ versus 0.3 to 0.5 MW/m/sup 3/), considerably reduced physical size of the fusion power core (FPC), and appreciably reduced economic leverage exerted by the FPC and associated physics result. The unique materials requirements anticipated for these compact reactors are outlined against the well documented ...

International Nuclear Information System (INIS)

... brakes lmfbr type reactors breeder reactors epithermal reactors fast reactors

Energy Technology Data Exchange (ETDEWEB)

Fission rate assessments are being made in FFTF to provide input for determination of power distribution, peak power, total power, fuel burnup distribution, conversion ration and neutron spectral information by the multiple foil reaction rate method. A goal of the initial FFTF Characterization Program is to measure isotopic fission rates to an accuracy of 2 to 5% throughout FFTF by radiometric techniques. To achieve this goal, measurements were made in the FFTF In-Reactor Thimble to reduce the uncertainties of key parameters needed for radiometric fission rate mapping of the FFTF core.

International Nuclear Information System (INIS)

Full text: Full text: The incomplete understanding of the complex mechanisms connected with the interaction between thermal-hydraulic and neutron kinetics still challenges the design and the operation of nuclear reactors and imposes the adoption of conservatism in the evaluation of safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience suggests the revisiting of those areas and the identification of design/operation requirements that can be relaxed. So far, almost all of the safety analyses of research reactors have been performed using conservative computational tools such as channel codes but, nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity. The global aim of the current work is an attempt to apply the best-estimate system thermal-hydraulic code Relap5. For this purpose, the generic IAEA ...

International Nuclear Information System (INIS)

There are several radioactive source terms in nuclear power plant's design and construction. The radioactivity source in systems and components is derived from the reactor coolant activity and provide the parameters used to determine secondary system equilibrium activities and annually releasing amounts to environment. The reactor coolant activity standard(ANSI-Nl8.l) had been periodically revised. In Korea, the utility should do the PSR for NPP's. The objective of PSR is to determine by means of a comprehensive assessment of an existing nuclear power plant to what extent the plant meets current internationally accepted safety standards and practices. So, Kori 3 NPP's reactor coolant activity is reviewing with the anticipated source terms. The comparative results of RCS average activity is lower one fifth (1/5) #approx# one tenth(1/10) than ANSI/ANS N18.1-1999.

International Nuclear Information System (INIS)

A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based ...

International Nuclear Information System (INIS)

As for No.2 plant in Sendai Nuclear Power Station, which is the fourth nuclear power generation facilities in Kyushu Electric Power Co., Inc., all works have been completed, and at present, the final trial operation is under way. In No.2 plant, many new techniques for raising the reliability and safety, improving the maintainability and reducing radiation exposure were introduced on the basis of the operation experience of PWRs obtained so far, similarly to No.1 plant. In this paper, the main items of the new techniques related to the design and construction of the plant are reported. No. 2 plant is a first improved and standardized plant having the thermal output of 2660 MW for standard three-loop PWRs, and the rated power output was set at 890 MW. As for the turbine, TC6F-40 in was adopted. As the improved design, a large reactor containment vessel, 17 x 17 type 9-grid fuel, ...

Energy Technology Data Exchange (ETDEWEB)

Forced vibration experiment has been conducted for the reactor building of Sendai Unit 1 nuclear power plant. The beam vibrational behaviors of the outer shielding building and the internal concrete structure have been observed by using a 50 tf vibration for low frequency region, and a 10 tf vibration for high frequency region, respectively. The outline of the experimental methods, the data handling system and the major results of experiment are described. The experimental results were simulated by an analytical model. The proper vibrational frequency and the vibration modes obtained by the analysis were compared with those obtained by the experiment. By these comparisons, the adequacy of the analytical method employed for the design was confirmed.

International Nuclear Information System (INIS)

Generally, Hardening and irradiated brominating phenomena are occurred in the reactor vessel under operation conditions by atomic cavities and creation of impurity atoms which are led by high fast neutron flux. To assure the mechanical integrity of pressure vessel until the end of power plant life after monitoring the sample specimens on the vessel inside, a series of tests is performed over the retrieved surveillance capsule to examine the changes according to the plant operation in accordance with regulations. Monitoring surveillance capsules attached to neutron shield wall of outer core are consists of impact sample, tensile sample and temperature monitor

International Nuclear Information System (INIS)

Several design options for improved passive circulation flow have been investigated for use in small, modular liquid metal cooled reactors (LMRs). The purpose is to enhance the transition to natural convection cooling following loss of forced circulation flow, reducing thermal transients experienced by the fuel and possibly eliminating the need for emergency pony-motor flow. Design details to minimize pressure drops may also enhance maximum equilibrium power levels possible under natural circulation only.

Energy Technology Data Exchange (ETDEWEB)

The performance of a solar chemical heat pipe was studied using CO{sub 2}reforming of methane as the endothermic reaction. A directly heated vertical reactor, packed with a rhodium catalyst was used. The solar tests were carried out in the Schaeffer solar furnace of the Weizmann Institute of Science. The power absorbed was up to 6.3 KW, the maximal flow rates of the gases reached 11,000 1/h, and the methane conversions reached 85%. A computer model was developed to simulate the process. Agreement of the calculations with the experimental results was quite satisfactory.

International Nuclear Information System (INIS)

Although there is service experience of Alloy 800 as tubing for superheaters in conventional and nuclear (HTR) power stations and in PWR heat exchangers, there is no corresponding service experience in sodium-cooled fast reactor steam generators. However, some limited experimental studies have been made of corrosion behaviour, and of possible structure modifications and effects on mechanical properties which occur during exposure of this material to a high temperature sodium environment, and these are summarised in the paper. It is concluded that further work needs to be done before Alloy 800 can be confidently endorsed for use as tubing in fast reactor steam generators. (author).

International Nuclear Information System (INIS)

This report presents results of the 1st regular inspection of the No.2 unit of the Sendai Nuclear Power Plant. It was carried out during the period from September 22, 1986, to December 24, 1986. The inspection covered the main unit of the nuclear reactor, facilities for the nuclear reactor cooling system, facilities for the instrumentation control system, fuel facilities, radiation control facilities, disposal facilities, nuclear reactor containment facilities, and emergency power generation system. Checking of appearance, disassemblage, leak and functions-performance of these facilities was conducted and no abnormalities were found. All operations involved in the inspection were performed under conditions within the permissible dose as specified in the applicable laws. No major modification work was carried out during the period of the regular inspection. The exposure dose ...

International Nuclear Information System (INIS)

Stretch-out operation mode can deepen the reactor burnup when the boron concentration is near 0 mg/L, in which the additional reactivity is introduced by the reducing of the moderator temperature and the decreasing of the load. Stretch-out is used in many nuclear power plants all over the world. The first stretch-out operation has been used for the first time in China. As a specific operation mode, which outruns the original reactor core design, the related and specialized design argument and safety analysis is required. As a consequence of the continuous or stepwise reduction of load and moderator temperature, the neurotic measurement system and the reactor control and protection system parameters should be modified specially. Based on the schedule of the electricity production, the first stretch-out operation had been carried out from March 12 to March 21 2003. It successfully avoided the overlapping ...

International Nuclear Information System (INIS)

The natural circulation experiments were conducted to confirm the cooling capability and the flow characteristics of the natural convection in the HANARO (Hi-flux Advanced Neutron Application Reactor). The tests were done at the power levels of 2%, 3% and 4% (1.2MW_t_h) of full power. The flow rates and temperatures at various locations of the primary and secondary cooling loops were measured at each power level. The temperature distributions in the chimney and the pool were also obtained. Through tests, the flow paths of the natural circulation and the cooling capability of the reactor were confirmed as designed. In addition, the simulation for the natural circulation tests was made by using RELAP5/KMRR, which was modified from RELAP5/MOD2 for applying to the HANARO conditions. The simulation results show that RELAP5/KMRR gives reasonable predictions for the flow rate and the ...

Energy Technology Data Exchange (ETDEWEB)

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making ...

International Nuclear Information System (INIS)

The formation and growth of silicon-nanoparticles from silane in a microwave reactor was investigated. Experiments were performed for the following conditions: precursor concentration 380-2530 ppm, pressures of 20-30 mbar, microwave powers 120-300 W. The formed particles were examined in-situ with a particle mass spectrometer. Additionally, particles were collected on grids and analyzed by transmission electron microscopy, X-ray diffraction, and by determining the specific surface area by BET. The particle size was found to be in the range of 5-8 nm in diameter. A simple model was used to simulate the particle formation processes taking place inside the reactor. The microwave energy coupled into the reactor flow was treated as a spatially distributed energy source resulting in a local temperature increase. The particles were assumed to have a monodisperse size distribution. To allow an approximation of ...

International Nuclear Information System (INIS)

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility has been in operation for approximately five years, including about one thousand days of full power operation of the Fast Test Reactor. During that time the collective dose equivalents received by operating personnel have been about two orders of magnitude lower than those typically received at commercial light water reactors. No major contamination problems have been encountered in operating and maintaining the plant, and release of radioactive gas to the environment has been minimal and well below acceptable limits. All shields have performed satisfactorily. Experience to date indicates an apparent radiological superiority of liquid metal reactor systems over current light water plants.

Energy Technology Data Exchange (ETDEWEB)

The renovation programme of the Phenix nuclear power plant (fast neutrons reactor situated at Marcoule) has for objective to ensure the reactor operation lengthening. In this frame, expertise and monitoring operation in situ of materials have been started. The presence of sodium and a temperature at the cold breakdown of the primary circuit between 150 and 180 degrees (Celsius) imply, for fast reactors, very special conditions. In this context, Framatome has realised three intervening in the area of nondestructive testing: the inspection of the cone-shaped support ring, the monitoring of the upper part of the primary vessel and the monitoring of the intermediary exchanger equipment. (N.C.)

International Nuclear Information System (INIS)

TAPP-3 and 4 reactors use large number of Self Powered Neutron Detectors (SPNDs) for Neutronic lower measurement and control. To perform in-situ calibration of these detectors in select locations and to validate the reactor physics codes which predict flux at various points in the core, traveling in-core probes (TIP) are required. The TIP assembly consists of a miniature neutron sensitive detector. The detector is driven in and out of core using a mechanism which facilitates positioning of the detector anywhere inside a vertical tube (Central carrier tube of any of the six select Vertical Flux Units) in the core. TIP is driven through retractable feed mechanism for a stroke of 13 m. This paper describes the developmental efforts and the operational feedback of the retractable feed mechanism for the stroke of 13 m used at TAPP 3 and 4 reactor. (author)

Energy Technology Data Exchange (ETDEWEB)

A comprehensive review was conducted of primary helium coolant chemistry data, based on current and past operating histories of helium-cooled, high-temperature reactors (HTGRs), including the Fort St. Vrain (FSV) HTGR. A reference observed FSV reactor coolant environment was identified. Further, a slightly drier expected FSV coolant chemistry was predicted for reactor operation at 100% of full power. The expected environment was compared with helium test environments used in the US, United Kingdom, Germany, France, and Japan. Based on a comprehensive review and analysis of mechanical property data reported for Alloy 800 tested in controlled-impurity helium environments (and in air when appropriate for comparison), an assessment was made of the effect of FSV expected helium chemistry on material properties of alloy 800, with emphasis on design properties of the Alloy 800 material utilized in the FSV ...

International Nuclear Information System (INIS)

Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top "1"6N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University ...

Science.gov (United States)

The theory of a method of determination of the reactivity from power spectral density measurements with /sup 252/Cf and the results of experiments with a critical assembly mock-up of a liquid-metal fast breeder reactor and with uranium (93.2 wt % /sup 235/U) metal cylinders and a sphere are presented. This method of reactivity determination has an advantage over existing methods in that it determines the reactivity only from properties of the reactor at the subcritical state of interest and thus does not require a calibration near delayed criticality. In these experiments the reactivity was varied by changing the fissile loading or the amount of neutron absorber inserted; for the LMFBR mock-up, the reactivity varied to approximately 75 dollars subcritical, and for the uranium metal assemblies, to approximately 30 dollars subcritical.

International Nuclear Information System (INIS)

The facilities in the Institute for Nuclear Research at Pitesti allow the testing, handling and examination of nuclear fuel and irradiated materials. The most important facilities are the TRIGA Steady State Research and Material Test Reactor and the Post-Irradiation Examination Laboratory (PIEL). The purpose of this work is to determine by post-irradiation examination, the behavior of CANDU fuel, irradiated in 14 MW TRIGA reactor. The fuel was irradiated in power ramp conditions. The results of post-irradiation examination are: - Visual inspection and photography of the outer appearance of sheath; - Profilometry (diameter, bending, ovality) and length measuring; - Determination of axial and radial distribution of the fusion products activity by gamma scanning and tomography; - Microstructural characterization by metallographic and ceramographic analyzes; - Mechanical properties determination. The data obtained from the ...

International Nuclear Information System (INIS)

The ROP margin in a CANDU reactor is decreasing over time due to the Primary Heat-Transport System (PHTS) aging effect. Adjustment of the ROP trip setpoint is required to maintain a high trip-probability and ROP trip effectiveness. Especially, for Wolsong-1, which is scheduled to change the old pressure tubes in 2009, the trend of ROPT after the retubing should be reevaluated. Before setting a ROPT, the main thermal characteristics including Critical Channel Power (CCP) should be calculated by the NUCIRC code. In this paper, the thermalhydraulic evaluation for Wolsung-1 was conducted with the updated Wolsung-1 PHTS data. Specifically, for the case of 0 EFPY (Effective Full Power Year) and 11 EFPY after the retubing, the distribution of the channel flow rate, channel exit quality, critical channel power, and critical power ratio (CPR) of the Wolsong-1 aged plant are calculated

Energy Technology Data Exchange (ETDEWEB)

Severe accident analysis has been performed for the Wolsung nuclear power plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks (LOAH) which is caused by loss of off-site power, diesel generators, and DC power, ISAAC(Integrated Severe Accident Analysis Code) computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanisms of calandria vessel and containment. In addition, some insights for accident management program (AMP) are also given.

International Nuclear Information System (INIS)

This paper reports on the Burning Plasma Experiment (BPX) which when operating at a toroidal field of 8.1 tesla and a plasma current of 10.6 megamps, requires peak power of 1235 megawatts and total pulse energy of over 21 gigajoules. These requirements are twice and over four times the corresponding figures for the Tokamak Fusion Test Reactor (TFTR), respectively. The design of the BPX power system has evolved, along with the tokamak, over a period of several years and has included studies of several alternative approaches. The reapplication of the existing TFTR power and energy facilities has been basic to all approaches. Among the new sources of pulse power and energy that have been considered are: direct utility grid pulsing, new flywheel units, and lead-acid storage batteries. The toroidal field power requirements are the greatest of the BPX subsystems and, ...

International Nuclear Information System (INIS)

This work estimates the power requirements for using photochemical processes driven by space nuclear power to counteract the Earth's ozone layer depletion. The total quantity of ozone (O_3) in the Earth's atmosphere is estimated to be about 4.7 x 10"3"7 molecules. The ozone production and destruction rates in the stratosphere are both on the order of 4.9 x 10"3"1 molecules/s, differing by a small fraction so that the net depletion rate is about 0.16 to 0.26% per year. The delivered optical power requirement for offsetting this depletion is estimated to be on the order of 3 GW. If the power were produced by satellite reactors at 800 km altitude (orbit decay time #approx# 300 years), some means of efficient power beaming would be needed to deliver the power to stratospheric levels (10--50 km). Ultraviolet radiation at 140--150 nm could have ...

International Nuclear Information System (INIS)

On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable ...

Energy Technology Data Exchange (ETDEWEB)

The world`s population of research reactors is growing old. Many have been adapted to serve new purposes over their lives, from testing materials for nuclear power programmes and supporting neutron physics experiments, to colouring gemstones, doping silicon and generating medical isotopes. In the first article of this survey of research reactor issues, Wilfried Krull from GKSS in Germany describes the effects on a reactor of supporting these changes in application as ``design ageing`` . Managing this and other symptoms of ageing to extend plant life is a key task for operators, and Krull discusses the efforts being made internationally to handle them. Eventually, terminal decline of one vital component can determine when a reactor has to be shutdown for refurbishment. For BR2 in Belgium, it was the beryllium matrix. Edgar Koonen from SCK-CEN explains work being done to replace it ...

Energy Technology Data Exchange (ETDEWEB)

A submerged nuclear power station has been proposed as an alternative power station with a relatively low environmental impact for use by both industrialized and developing countries. The station would be placed 10 m above the seabed at a depth of 30--100 m and a distance of 10--30 km from shore. The submerged nuclear power station would be manufactured and refueled in a central facility, thus gaining the economies of factoryfabrication and the flexibility of short-lead-time deployment. To minimize the size of the submerged hull, horizontal steam generators are proposed for the primary-to-secondary heat transfer, instead of the more traditional vertical steam generators. The horizontal steam generators for SNPS would be similar in design to the horizontal steam generators used in the N-Reactors except the tube orientation is horizontal (the tube`s inlet and outlet connection points on the tubesheet are ...

International Nuclear Information System (INIS)

The generation of fusion power from the Fast-Liner Reactor (FLR) concept envisages the implosion of a thin (3-mm) metallic cylinder (0.2-m radius by 0.2-m length) onto a preinjected plasma. This plasma would be heated to thermonuclear temperatures by adiabatic compression, pressure confinement would be provided by the liner inertia, and thermal insulation of the wall-confined plasma would be established by an embedded azimuthal magnetic field. A 2- to 3-mu s burn would follow the approx. 10"4 m/s radial implosion and would result in a thermonuclear yield equal to 10 to 15 times the energy initially invested into the liner kinetic energy. For implosions occurring once every 10 s a gross thermal power of 430 MWt would be generated. The results of a comprehensive systems study of both physics and technology (economics) optima are presented. Despite unresolved problems associated with both the physics and technology of the FLR, ...

International Nuclear Information System (INIS)

... computer calculations fftf reactor nonlinear problems reactor accidents reactor

International Nuclear Information System (INIS)

The author gives the historical development of steam-turbine construction in Europe since the turn of the century, and the technical further development of conventional turbines due to the increases in the steam parameters and per-unit outputs in the increases in the steam parameters and per-unit outputs in Europe and the USA. Marginal conditions for the development of turbines in nuclear power stations with light-water reactors are mentioned. The rise in the per-unit capacities of the turbosets constructed in Germany and the USA for nuclear power stations is discussed. Longitudinal sections through typical turbines are shown. The future development of turbines with high output is dealt with. (orig.).

International Nuclear Information System (INIS)

In the last 30 years the production of electricity in the USSR has increased 14-fold, probably attaining 1540 billion kWH in 1985. Nuclear generation will provide the bulk of future increases of consumption, using both water-cooled and uranium/graphite reactors; stations of up to 1.5 million kW are in service. The USSR is also in the fore-front of attempts to exploit thermonuclear power. The USSR is also conducting experiments with renewable sources of energy such as solar, geothermal, wind and wave power and with magnetohydrodynamic generation. (D.A.J.).

International Nuclear Information System (INIS)

This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned ...

International Nuclear Information System (INIS)

The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort ...

International Nuclear Information System (INIS)

By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the ...

Energy Technology Data Exchange (ETDEWEB)

The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power Plants.`` The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple ...

International Nuclear Information System (INIS)

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety of the facility. The three passive safety attributes being evaluated in the paper are: 1) In-core and in-vessel natural convection cooling, 2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and 3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond to most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) models ...

Energy Technology Data Exchange (ETDEWEB)

The {sup 252}Cf-source-driven noise analysis method is a versatile measurements tool that has been applied to measurements for initial loading of reactors, quality assurance of reactor fuel elements, fuel processing facilities, fuel reprocessing facilities, fuel storage facilities, zero-power testing of reactors, verification of calculational methods, process monitoring, characterization of storage vaults, and nuclear weapons identification. This method`s broad range of application is due to the wide variety of time- and frequency domain signatures, each with unique properties, obtained from the measurement. The following parameters are obtained from this measurement: average detector count rates, detector multiplicities, detector autocorrelations, cross-correlation between detectors, detector autopower spectral densities, cross-power spectral densities between detectors, ...

International Nuclear Information System (INIS)

First structural wall material (FSW) materials in inertial fusion energy (IFE) power reactors will be irradiated under typical repetition rates of 1-10 Hz, for an operation time as long as the total reactor lifetime. The main objective of the present work is to determine whether a continuous-pulsed (CP) approach can be an efficient method in modeling the pulsed activation process for operating conditions of FSW materials. The accuracy and practicability of this method was investigated both analytically and (for reaction/decay chains of two and three nuclides) by computational simulation. It was found that CP modeling is an accurate and practical method for calculating the neutron-activation of FSW materials. Its use is recommended instead of the equivalent steady-state method or the exact pulsed modeling. Moreover, the applicability of this method to components of an IFE power plant subject to ...

Science.gov (United States)

Chemical looping combustion (CLC) is a process in which an oxygen carrier is used for fuel combustion instead of air or pure oxygen as shown in the figure below. The combustion is split into air and fuel reactors where the oxidation of the oxygen carrier and the reduction of the oxidized metal occur respectively. The CLC system provides a sequestration-ready CO2 stream with no additional energy required for separation. This major advantage places combustion looping at the leading edge of a possible shift in strict control of CO2 emissions from power plants. Research in this novel technology has been focused in three distinct areas: techno-economic evaluations, integration of the system into power plant concepts, and experimental development of oxygen carrier metals such as Fe, Ni, Mn, Cu, and Ca. Our recent thorough literature review shows that multiphase fluid dynamics modeling for CLC is not available in the ...

International Nuclear Information System (INIS)

Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

Energy Technology Data Exchange (ETDEWEB)

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

International Nuclear Information System (INIS)

The radiation life time of reactor pressure vessel (RPV) is the most important limiting factor for the term of exploitation of the whole power unit. The main degradation mechanism of RPV metal is the neutron induced embrittlement. Processes of radiation ageing running in RPV metal lead to fracture toughness decrease and to increased probability of brittle fracture of the vessel under thermal shocks. This explains the importance of RPV integrity assessment and rest life time management

International Nuclear Information System (INIS)

A quadruply redundant synchronous fault tolerant processor (FTP) is now under fabrication at the C.S. Draper Laboratory to be used initially as a trip monitor for the Experimental Breeder Reactor EBR-II operated by the Argonne National Laboratory in Idaho Falls, Idaho. The real time operating system for this processor is described.

International Nuclear Information System (INIS)

Appendix B to 10 CFR Part 50 establishes overall quality assurance requirements for the design, construction and operation of safety-related structures, systems, and components. This guide presents a method acceptable to the Commission for complying with these regulations with regard to overall quality assurance program requirements during design and construction of nuclear power plants. Input to this guide has been provided by the Advisory Committee on Reactor Safeguards.

Energy Technology Data Exchange (ETDEWEB)

A krypton recovery pilot plant has been completed for the Power Reactor and Nuclear Fuel Development Corporation. This is the first industrial facility in the world to make practical use of development results for offgas treatment and storage from nuclear facilities. The cryogenic distillation process was adopted as a proven and reliable method to separate krypton and xenon, and to reduce gas effuents to a level so low that the decontamination factor amounts to more than 1000.

International Nuclear Information System (INIS)

The Flux Mapping System (FMS) of 540 MWe PHWR is a system, which is first of its kind used in Indian PHWRs. It is used to compute a detailed flux/power distribution of the reactor core using modal synthesis method .The paper brings out the high availability features of FMS and the software design philosophy. The paper emphasizes on framework based reusable architectural design, which simplifies and speeds up the development of data acquisition systems. (author)

International Nuclear Information System (INIS)

The Fast Flux Test Facility on the federal reservation at Hanford, Washington, has become a bright star in the universe of nuclear science and engineering technologies. The entire FFTF enterprise is now a success story, and this is particularly significant in these days when good news about nuclear power is scarce. The reactor, its testing capabilities and associated test facilities are described.

Energy Technology Data Exchange (ETDEWEB)

The technical rule intends to maintain the function of the safety facilities in case of explosion hazards resulting from materials capable of generating explosive atmospheres or forming explosive mixtures if such materials are brought or released into the nuclear facility or are generated on site. (orig.).

International Nuclear Information System (INIS)

In order to implement effective safety regulations on aging management for reactor facilities etc., the information on important technology issues, the latest technical knowledge including evaluation technology, test and research outcomes, related codes and standards, regulation information, operation experiences such as accidents and trouble, etc. with respect to aging-induced deterioration in and outside Japan and in other industries, were collected, organized and evaluated. (author)

International Nuclear Information System (INIS)

Large scale flows of liquids can be controlled by using power fluidic devices that harness the hydrodynamic properties of liquids rather than use moving parts. Included among the fluidic devices considered are fluidic pumps, reverse flow diverters, fluidic diodes and vortex amplifiers. These devices are of potential use in the nuclear industry, particularly in reprocessing. (U.K.).

International Nuclear Information System (INIS)

The Nuclear Safety Commission acknowledged the policy of the Agency of Natural Resources and Energy to cope with these important matters on August 8, 1984. The main contents of the investigation and deliberation were as follows, As to the prevention of the occurrence of fires, the prevention of the leak and spread of inflammable liquid and gas, the installation of protective relays and so on, the use of incombustible materials and aseismatic design and the installation of lightning arresters. As to the detection of fires and fire fighting, the proper selection and arrangement of fire detectors and extinguishers, the extinguishers which do not harm the safety function of structures and equipment, and the extinguishers which are not affected by natural phenomena. As to the reduction of the effect of fires, the proper installation of fire walls and extinguishers, and the high temperature shut off of nuclear reactors which is never hampered by any fire. (Kako, I.).

International Nuclear Information System (INIS)

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a ...

International Nuclear Information System (INIS)

This paper presents actual time/motion data for an oversize truck spent-fuel shipment from its origin, Surry, Virginia to its destination, Idaho National Engineering Laboratory (INEL). These data include the receipt of the empty cask at the reactor, wet-loading the cask, over-the-road or in-transit data, and receipt and dry unloading of the shipping cask at the receiving facility. Occupational doses were recorded at the Surry Power Plant as well as at INEL, and public doses were calculated for the in-transit dose analysis. This shipment was one of a series performed in support of a demonstration and evaluation of dry storage at INEL. The oversized shipment consisted of a TN-8L shipping cask loaded with three 10-yr-old pressurized water reactor assemblies. The total distance traveled was #approx#2800 miles, requiring 62 h including stops. The time required to receive and inspect the empty shipping cask and wet-load and ...

International Nuclear Information System (INIS)

A calculation program (URA 6.F4) was elaborated on FORTRAN IV language, that through finite differences solves the unidimensional scalar Helmholtz equation, assuming only one energy group, in spherical cylindrical or plane geometry. The purpose is the determination of the flow distribution in a reactor of spherical cylindrical or plane geometry and the critical dimensions. Feeding as entrance datas to the program the geometry, diffusion coefficients and macroscopic transversals cross sections of absorption and fission for each region. The differential diffusion equation is converted with its boundary conditions, to one system of homogeneous algebraic linear equations using the box integration technique. The investigation on criticality is converted then in a succession of eigenvalue problems for the critical eigenvalue. In general, only is necessary to solve the first eigenvalue and its corresponding eigenvector, employing the power method. The ...

International Nuclear Information System (INIS)

The Swedish State Power Board has together with Nukem, Hanau, West-Germany carried out pyrolysis o powder resins in a pilot plant with a capacity of about 30 kg/hr. The pyrolysis reactor with its afterburner and offgas scrubber system has been operated under steady state condition. About 2200 kg resins have been pyrolysed under November-December 1983 and the decontamination factor for Cs has been measured. Solidification of the residues from the pyrolysis reactor and scrubber water solutions has been carried out and various recipes with cement have been tested. The pyrolysis process has high decontamination factors and no offgas problems as the operating temperature of the reactor is Low. The residues from the reactor are chemically dead and can not cause swelling problems. Compared with a normal cementation process the final waste volume will be reduced with a factor of 4 if also ...

International Nuclear Information System (INIS)

In the radiochemical reprocessing of spent fuel from nuclear reactors, a considerable amount of liquid, solid, and gaseous waste is formed; this waste is potentially dangerous to humans and requires the development of special and complex technological techniques for its localization and reliable long-term storage. The most hazardous are liquid wastes of high specific activity - water-tailings solutions obtained in the first cycle of extraction after the removal of uranium and plutonium. These solutions contain more than 99.9% of all the other transuranic elements - isotopes of neptunium, americium, and curium. Where necessary, some fission products and actinides may be removed from wastes of high specific activity for subsequent use. The quantity, composition, and activity of these wastes varies within broad limits, depending on the type and power of the reactor, the initial nuclide composition of the fuel, and its specific ...

International Nuclear Information System (INIS)

The Surry low power and shutdown probabilistic risk analysis (PRA) is an ongoing project at Brookhaven National Laboratory (BNL) to identify and quantify potential accident scenarios that may occur in a pressurized water reactor (PWR) during low power and shutdown. It was initiated as a result of various incidents and accidents that have occurred within the United States and overseas. The project involves review and evaluation of PWR experience at shutdown, identification of accident scenarios, determination of methods to mitigate the accidents, and performance a level 1 PRA. An evaluation of accident progression, source terms and consequences has also been initiated. The results will be used to address issues related to shutdown conditions. The objective of this paper is to provide a progress report on the project, and to present the approach used as well as the preliminary results of the ongoing and completed tasks.

Energy Technology Data Exchange (ETDEWEB)

Supplement 2 to the Safety Evaluation Report for Mississippi Power and Light Company, et. al, joint application for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson, in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.

Energy Technology Data Exchange (ETDEWEB)

The prospects for the low-aspect-ratio (A) tokamak to fulfill the requirements of viable fusion power plants are considered relative to the present status in data and modeling. Desirable physics and design features for an attractive Blanket Test Facility and power reactors are estimated for low-A tokamaks based on calculations improved with the latest data from small pioneering experiments. While these experiments have confirmed some of the recent predictions for low-A, they also identify the remaining issues that require verification before reliable projections can be made for these deuterium-tritium applications. The results show that the low-A regime of small size, modest field, and high current offers a path complementary to the standard and high A tokamaks in developing the full potential of fusion power.

Energy Technology Data Exchange (ETDEWEB)

Fuel elements used in the ORR whole-core LEU fuel demonstration have been gamma-scanned to determine axial distributions of UZLa and TXCs fission product activities. This data has been analyzed to determine cycle-averaged fuel element powers, residual STVU masses, and burnups of discharged fuel elements. Methods used to analyze the data are discussed and results are presented for the LEU fuel elements. Measured and calculates fuel element powers agree to within 5%, residual STVU masses to within 2%, and burnups to within 3%. These results are somewhat preliminary and await improved burnup calculations and independent calibration data to be based on the destructive analyses of a number of irradiated fuel elements. 4 refs., 7 figs., 3 tabs.

International Nuclear Information System (INIS)

Novel and unprecedented ion exchange resin cleaning system, for use in BWR plants and featuring a vibration separator and basic design factors of Radiological Solutions, Inc., had been delivered to Tokai No. 2 Power Station, Japan Atomic Power Company, in October 2005. This compactly-designed system effectively separates crud and resin fines from ion exchange resins, with no clogging of separation screens. It generates minimized waste liquid and has a specially designed over-pack cleaning tank. The system has been in operation for about 2 years and half now and favorable operational data, such as crud and sulfate concentration decrease in feed water and reactor water respectively, and evaluation results have been reported from Japan Atomic Power Company and so on. (author)

International Nuclear Information System (INIS)

Commissioning of a Pressurized heavy water reactor (PHWR) plant of NPCIL involves three phases viz phase-A which consist of pre-criticality activities such as hydro test, air hold test, no load test of motors etc., phase-B consist of criticality and post criticality physics experiments. The phase-C, which is considered the major phase, consist of initial power raise to about 10 % , TG rolling, synchronization, going to significant power in steps and performance tests such as load rejection tests from various power levels. In order to have smooth commissioning for the Phase-C, an integrated team consisting of engineers from various design and analysis groups of NPCIL headquarters was formed to participate along with site O and M engineers, closely observe and coordinate phase-C commissioning activities. During this commissioning some major events and observations took place. An attempt is made to bring ...

Energy Technology Data Exchange (ETDEWEB)

The NUclear Power Engineering Corporation (NUPEC) has developed IMPACT-SAMPSON code to analyze integral behavior of light water nuclear power plants under severe accident conditions. IMPACT-SAMPSON's distinguishing features include interconnected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to severe accident events, and high-speed simulation on parallel processing computers. The integral plant behaviors of typical PWR and BWR under severe accident conditions have been analyzed with the IMPACT-SAMPSON code. The PWR plant analyzed was the three-loop, steel-dry containment type with 2,440 MWt. The AE accident scenario was supposed, that is, LOCA by 6-inch hot leg failure followed by accumulated water injection, but no ECCS and containment spray activation. The BWR plant analyzed was the 3,293 MWt BWR-5, Mark-II containment type. The TQUV accident scenario was supposed, ...

International Nuclear Information System (INIS)

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor ...

International Nuclear Information System (INIS)

One of the main points for the inherent safety of a pebble bed high temperature reactor (HTR) is to guarantee the safe removal of the after-heat in case of a break-down of all active cooling systems like heat-exchangers or liner-cooling. This will be necessary because it is well known today that graphite pebble bed fuel elements stay intact, if the accident temperature is below 1600 deg. C. Therefore the heat must be taken out of the reactor system by passive, natural law heat-transfer mechanism so that the maximum fuel temperature stays below the specified limit. Today medium-sized HTRs with a power of 750 MW_t_h and more (TGTR-300, HTR 500) reach temperatures of more than 2400 deg. C in small parts of the core in such hypothetical accidents. A possible way to realize the inherent safe heat removal in advanced medium-sized HTRs is to change the form of the core. Instead of employing the standard cylindrical geometry a ...

International Nuclear Information System (INIS)

Large-scale decommissioning of Russian nuclear-powered submarines (NPS) and their utilization prospects gave rise to numerous complicated scientific and technical, as well as economic, problems. Problems of handling of radioactive equipment from the reactor compartments (RC) are among the vital ones, arousing a growing concern with the public. Without solution of the problems the processes of NPS utilization can not be considered completed. It involves potential hazard, for the environment both from NPS being paid up (temporal on-float storage) with unloaded spent nuclear fuel (SNF), and RC, cut from submarine hull, containing highly radioactive equipment and materials but no SNF. Diverse variations of the concept of reactor compartment handling of NPS subject to, utilization are possible, but, in principle, there are essentially two variants: (1) RC utilization directly in the course of NPS utilization, envisaging removal ...

Energy Technology Data Exchange (ETDEWEB)

Four 10 MW, highly-stabilized power supply modules have been installed at the National High Magnetic Field Laboratory in Tallahassee, FL, to energize water-cooled, resistive, high-field research magnets. The power supply modules achieve a long term current stability if 10 ppM over a 12 h period with a short term ripple and noise variation of <10 ppM over a time period of one cycle. The power supply modules can operate independently, feeding four separate magnets, or two, three or four modules can operate in parallel. Each power supply module consists of a 12.5 kV vacuum circuit breaker, two three-winding, step-down transformers, a 24-pulse rectifier with interphase reactors, and a passive and an active filter. Two different transformer tap settings allow rated dc supply output voltages of 400 and 500 V. The rated current of a supply module is 17 kA and each supply module has a ...

Energy Technology Data Exchange (ETDEWEB)

An almost primordial trend in the conversion and use of energy is an increased complexity and cost of conversion systems designed to utilize cheaper and more-abundant fuels; this trend is exemplified by the progression fossil fission {yields} fusion. The present projections of the latter indicate that capital costs of the fusion ``burner`` far exceed any commensurate savings associated with the cheapest and most-abundant of fuels. These projections suggest competitive fusion power only if internal costs associate with the use of fossil or fission fuels emerge to make them either uneconomic, unacceptable, or both with respect to expensive fusion systems. This ``implementation-by-default`` plan for fusion is re-examined by identifying in general terms fusion power-plant embodiments that might compete favorably under conditions where internal costs (both economic and environmental) of fossil and/or fission are not as great as is needed to justify ...

International Nuclear Information System (INIS)

Forced vibration experiment has been conducted for the reactor building of Sendai Unit 1 nuclear power plant. The beam vibrational behaviors of the outer shielding building and the internal concrete structure have been observed by using a 50 tf vibration for low frequency region, and a 10 tf vibration for high frequency region, respectively. The outline of the experimental methods, the data handling system and the major results of experiment are described. The experimental results were simulated by an analytical model. The proper vibrational frequency and the vibration modes obtained by the analysis were compared with those obtained by the experiment. By these comparison, the adequacy of the analytical method employed for the design was confirmed. (Aoki, K.).

International Nuclear Information System (INIS)

Modern civilisation is based on substantial utilisation of energy. Rapid industrial development and improvement of living standards in India require energy planners to adequately forecast the energy demand and take appropriate measures in advance. However, the development and establishment of new technology is a slow process, sometimes extending over decades. Hence, energy options based on new technologies need to be planned for much in advance making allowance for uncertainties and delays. Fast Breeder Reactor (FBR) technology is an advanced energy option promising abundant and economic supply of power. Research and development work on FBRs has been conducted at the Indira Gandhi Centre of Atomic Research (IGC) since 1971. The international trends in FBR development are highlighted in this discussion and an overview of some of the research activities at IGC is presented. (author). 8 refs., 7 tab s.

International Nuclear Information System (INIS)

For shut-down safety of the turbine generator (securing of auxiliary power operation after load shut-down and preventing the reaching of overspeed after load shut-down with disturbed turbine governing system) additional measures compared to those for superheated steam turbines are required for turbine generators in plants with pressurized water reactor (PWR) as well as those with boiling water reactor (BWR) . Equipment is described (e.g. overspeed govern or selecting connection, vacuum breaker, bypass valves, intercepting valves) which, depending on the own conditions of the individual turbine generator (e.g. run-up time, vacuum, enclosed energy), may be applied alone or in jointly. (orig.).

International Nuclear Information System (INIS)

This paper will adress the unique safety related aspects of the LMFBR concept which are of significance to containment design and structural analyses. Topics to be covered will include: primary boundary integrity assurance; effects of sodium spills on integrity of structures; provisions being considered for containment of melted cores; and fuel handling accidents. Specific reference will be made to the FFTF and the Clinch River Breeder Reactor Project designs and methods of treatment of the above problems. In particular, the part played by tests, such as those carried out on a simulated FFTF model, and the planned structural reliability and related programs will be considered. Where practicable, these topics will be addressed in a manner which places FFTF and CRBR in context with other LMFBR's, and will point to a possible direction for future American LMFBR designs.

International Nuclear Information System (INIS)

Ppercase(femaxi-iv) is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of ppercase(femaxi-iv) is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of ppercase(femaxi-iv) was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in ppercase(femaxi-iv) and some results of applications on the experimental data. ((orig.)).

International Nuclear Information System (INIS)

Reactor Control Division (RCnD) has been one of the main designers of safety and safety related systems for power reactors. These systems have been built using in-house developed hardware. Since the present set of hardware was designed long ago, a need was felt to design a new family of hardware boards. A Working Group on Electronics Hardware Standardization (WG-EHS) was formed with an objective to develop a family of boards, which is general purpose enough to meet the requirements of the system designers/end users. RCnD undertook the responsibility of design, fabrication and testing of boards for embedded systems. VME and a proprietary I/O bus were selected as the two system buses. The boards have been designed based on present day technology and components. The intelligence of these boards has been implemented on FPGA/CPLD using VHDL. This paper outlines the various boards that have been developed with a brief ...

Energy Technology Data Exchange (ETDEWEB)

Femaxi-iv is a computer code for the analysis of thermal and mechanical behavior of light water reactor fuel rods during steady-state and transient conditions. The main purpose of femaxi-iv is to calculate the stress and strain distributions in the fuel and cladding due to the pellet-cladding mechanical interaction, and the fission gas release rate during operations, especially power transients. The capability of femaxi-iv was extensively tested with a number of experimental results obtained in some international fuel irradiation programs. This paper provides a general description of the various models involved in femaxi-iv and some results of applications on the experimental data. ((orig.))

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400 Mwt sodium-cooled fast reactor operating at the Hanford Engineering Development Laboratory, Richland, Washington, to conduct fuels and materials testing in support of the US Liquid Metal Fast Breeder Reactor (LMFBR) program. Startup and initial power testing included a comprehensive series of nonnuclear and nuclear tests to verify the thermal, hydraulic, and neutronic characteristics of the plant. A specially designed series of natural circulation tests were then performed to demonstrate the inherent safety features of the plant. Early in 1982, the FFTF began its first 100-day irradiation cycle. Since that time the plant has operated very well, achieving a cycle capacity factor of 94% in the most recent irradiation cycle. Seventy-five specific test assemblies and 25,000 individual fuel pins have been irradiated, some in excess of 80 MWd/Kg.

International Nuclear Information System (INIS)

In Tarapur Atomic Power Station unit-3 and 4, which is 540 MWe pressurized heavy water reactor, tritium is produced in primary heat transport system and moderator system. Tritium is a major contributor to the internal dose. Internal dose contributes about 30% of the collective dose. Internal dose monitoring and its control are important to control the collective dose. Estimation of internal dose is done by analysis of bioassay samples of radiation workers. In a month, about 7000 bioassay samples are analysed for the internal dose assessment during normal operation, and about 12000 during the biennial shut down of the reactor. To enhance the sample preparation and counting performance, minimize the entry errors and reduce the processing time, barcode based label generation system was developed for the internal dose monitoring. This paper discusses about the use of barcode system in the internal dose monitoring at TAPS 3 and ...

Energy Technology Data Exchange (ETDEWEB)

Recently, the leak-before-break (LBB) concept has been used successfully to eliminate some pipe whip restraints, snubbers and jet impingement shields from the primary reactor cooling system piping of pressurized water reactors. This has resulted in substantial savings in maintenance costs, reductions in radiation exposure of plant service personnel, and has enhanced the overall safety of nuclear power plants. This study provides guidelines to utilities in expanding the application of the LBB concept to additional pipe systems and it couples the concept with hardware optimization. Seven high energy piping systems were investigated for technical feasibility in using the LBB concept. The results indicate that some of these seven lines are good candidates for the leak-before-break application.

International Nuclear Information System (INIS)

The AVR experimental nuclear reactor was Professor Dr. Rudolf Schulten's brainchild. Visionary ideas led to the success of this technology: - Helium coolant because of the particularly high heat transfer coefficients; - an integrated primary system reactor concept as the basis of all safety considerations in the interest of maximum safety; - uranium-235 and thorium-232 fuel allowing new fuel to be bred; - high temperatures for electricity generation at maximum thermodynamic efficiencies, i.e. optimum fuel utilization; - the possibility to run chemical processes economically at high temperatures by means of nuclear fuels; - the inherent safety of the reactor, for a major accident accompanied by a complete loss of cooling cannot occur for nuclear physics reasons, as was tested twice in the AVR. The AVR attained its first criticality on August 28, 1968. It was operated for more than 20 years, until December 31, 1988, at ...

Energy Technology Data Exchange (ETDEWEB)

In the fall of 2003, the operators noticed that in the recently-refurbished Bruce A Shutdown System no. 1 (SDS1) the noise level in Log Rate signals were much larger than before. At the request of the Canadian Nuclear Safety Commission (CNSC), all Canadian CANDU reactors took action to characterize their Log Rate noise. Staff of the Inspection and Maintenance Services division of Ontario Power Generation (OPG) has collected high-speed high-accuracy noise data from nearly all 16 Ontario reactors, either as part of routine measurements before planned outages or as a dedicated noise recording. This paper gives the results of examining a suitable subset of this data, with respect to the characteristics and possible causes of Log Rate noise. The reactor and instrumentation design is different at each station: the locations of the moderator injection nozzles, the location of the ion chambers for each system, ...

International Nuclear Information System (INIS)

An experimental program has been conducted to determine the feasibility of natural convection cooling of a reactor following a beyond-design-based accident. The particular application under consideration was the heavy-water new production reactor. The questions to be resolved include the verification that a natural convection cooling pattern would be established and the determination of the power limit for which convective cooling will occur for a significant period of time. In the experiment, the reactor configuration was simulated using small-diameter vertical heated tubes in parallel with a large-diameter bypass line. Following a loss-of-flow event, the flow in the bypass line will reverse direction and pass through the heated channel by means of natural convection. If, however, the channel power is too high, void formation will block the channel and prevent the reverse flow ...

International Nuclear Information System (INIS)

CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO_2), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An ...

Energy Technology Data Exchange (ETDEWEB)

On March 28, 1979, the Pennsylvania Three Mile Island nuclear power plant Unit 2 (TMI-2) suffered a partial fuel-melt accident. During this accident, over 20,000 lb of molten fuel flowed through holes melted through the baffle plates and through the lower-core support assembly (LCSA). The molten fuel subsequently resolidified in the bottom of the reactor vessel. The lower-core support assembly of the TMI-2 reactor was not structurally damaged during the accident. In order to permit defueling of that region of the core, the LCSA was cut to permit access. A five-axis teleoperator was developed to deliver plasma arc cutting, rotary grinding and abrasive waterjet cutting of end effectors to the LCSA. Complex geometry sectioning was completed in a mock-up facility at chemistry and pressure conditions simulating those of the vessel, prior to actual in-vessel operations. In-vessel activities began in early May 1988 and were ...

International Nuclear Information System (INIS)

The RADionuclide Transport, Removal, And Dose (RAD-TRAD) code is designed for U.S. Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the off-site population and to control room operators following a design-basis accident at light water reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465. The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken, including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available for these different removal mechanisms that use, as input, ...

International Nuclear Information System (INIS)

The authors present a state-of-the-art review of natural circulation heat transfer in loop type reactor plants. Most of the examples are taken from Fast Flux Test Facility (FFTF) design experience, drawing on the authors' familiarity and a developing base of available documentation. On-going studies related to the Clinch River Breeder Reactor (CRBR) and some foreign experience are also noted where available in the literature. The emphasis is on the role of natural circulation in decay heat removal; however, free convection during either operation at power or normal shutdown does influence some aspects of the design and these are reviewed. In treating decay heat removal the topics discussed include steady state loop performance and transient dynamics for conditions immediately after scram and for the longer term which involves different considerations. The review summarizes complex dynamics, specific to the FFTF design ...

Energy Technology Data Exchange (ETDEWEB)

Activities are reported by the Reactor Systems Section, Research Instrument Section, and the Measurement and Controls Engineering Section. Reactor system activities include dynamic analysis, survillanc and diagnostic methods, design and evaluation, detectors, facilities support, process instrumentation development, and special assignments. Activities in the Research Instrument Section include the Navy-ORNL RADIAC development program, advanced ..gamma.. and x ray detector systems, neutron detection and subcriticality measurements, circuit development, position-sensitive detectors, stand-alone computers, environmental monitoring-detectors and systems, plant security, engineering support for fusion energy division, engineering support for accelerator physics, and communications: radio, closed-circuit tv, and computer. Activities in the Measurement and Controls Engineering Section include the AVLIS program; gas centrifuge enrichment technology ...

Science.gov (United States)

Environmental characteristics of conceptual fusion-reactor systems based on magnetic confinement are examined quantitatively, and some comparisons with fission systems are made. Fusion, like all other energy sources, will not be completely free of environmental liabilities, but the most obvious of these-- tritium leakage and activation of structural materials by neutron bombardment-- are susceptible to significant reduction by ingenuity in choice of materials and design. Large fusion reactors can probably be designed so that worst-case releases of radioactivity owing to accident or sabotage would produce no prompt fatalities in the public. A world energy economy relying heavily on fusion could make heavy demands on scarce nonfuel materials, a topic deserving further attention. Fusion's potential environmental advantages are not entirely ''automatic'', converting them into ...