Science.gov (United States)

In order to reduce the air concentration of (sup 3)H in the reactor buiIding of Wolsung Heavy Water Reactor, a computer code for estimation of adsorption behavior was programmed based on an equation derived for analysis of water vapor adsorption, and a ba...

Energy Technology Data Exchange (ETDEWEB)

For light water reactors, loss of coolant is an important point in safety analysis, whereas for gas-cooled reactors the ingress of water into the core region is an incident of safety relevance. The applicability of the computer code system GAMTEREX to pebble beds of spherical high-temperature gas-cooled reactor fuel elements with simulated water ingress is verified by experiment. The measurements were performed at a Siemens-Argonaut reactor, using its ring core as a driver zone for a pebble-bed core in the center of the reactor.

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.

Energy Technology Data Exchange (ETDEWEB)

The ability of the /sup 252/Cf-source-driven neutron noise analysis method to measure subcriticality has been demonstrated in a variety of experimental configurations of fissile materials. Calculations for an approximately 4-m-dia configuration of light water reactor (LWR) fuel elements indicated the feasibility of measuring the subcriticality of large, loosely coupled arrays of LWR fuel elements by this same method. These analysis suggested application to the initial loading of both pressurized and boiling water reactors, zero-power testing of reactors (such as shutdown margin measurements after initial loading), light water reactor refueling, and safe storage of LWR spent fuel. In the fuel storage application, direct measurement of subcriticality in the actual fuel storage facilities provides the ...

International Nuclear Information System (INIS)

RELAP5 (Reactor Excursion and Leak Analysis Program) is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor UMLRR are used. The UMLRR is a 1 MW, light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the ...

International Nuclear Information System (INIS)

SCDAP/RELAP5 code has been developed in US for best-estimate simulation of light water reactors transients during nuclear accidents. The code models the coupled behaviour of the cooling system, reactor core and fission products release during the accident. It is the result of the coupling between RELAP5, modelling thermal hydraulic, control system, reactor kinetics and the transport of noncondensable gases, and SCDAP code modelling the behaviour of the reactor core during severe accidents. The paper briefly presents the application of SCDAP/RELAP5 code to CANDU severe accident analysis. Also, the paper proposes a summary of the needs for development that could enhance the quality of the severe accidents related predictions in CANDU reactors. (authors)

Energy Technology Data Exchange (ETDEWEB)

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of ...

International Nuclear Information System (INIS)

On the 3. and 4. November 1982 the sixth conference of the Corporation for Reactor Safety (GRS) was held in Cologne's Guerzenich. The theme of this year's meeting was the 'Status of Risk Investigations at Nuclear Power Plants'. A principal topic was a report on findings made by the GRS during the 'Risk Oriented Analysis SNR-300'. The second topic comprised the newest developments within Phase B of the Risk Study of Water Pressure Reactors, the discussion of the dose/effect relationship and considerations on threshold risk values. (orig.).

International Nuclear Information System (INIS)

Research highlights: ? We model power oscillations in boiling water reactors using a lumped parameter model. ? The nature and amplitudes of oscillations is obtained using a nonlinear analysis. ? The method of multiple scales has been used for the analytical treatment. ? Fuel temperature coefficient of reactivity determines the nature of oscillations. ? The presented systematic method of analysis useful for reduced order reactor models. - Abstract: In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy ...

International Nuclear Information System (INIS)

Level 1 and 2 probabilistic safety assessments (PSAs) for both internal and external events are being performed to meet one of the conditions for a construction permit for Wolsong units 2, 3, and 4 in Korea. These units are CANDU-6 Pressurized Heavy Water Reactors (PHWRs), and the study is the first comprehensive level 1 and 2 PSAs for CANDU type plants in the world. The detailed PSA includes and extensive fault tree, event tree analysis, human reliability analysis, and common cause failure analysis. Event trees have been developed for 35 internal initiating event groups. The preliminary results show that the total core damage frequency for Wolsong units 2, 3, and 4 each is similar to that for a typical PWR plant. (author).

Energy Technology Data Exchange (ETDEWEB)

The core bypass phenomenon of borated water injected through direct vessel injection (DVI) nozzles in APR1400 (Advanced Power Reactor 1400MWe) during main steam line break (MSLB) accidents with a reactor coolant pump (RCP) running mode has been simulated using a two-channel and one-dimensional system analysis model code (MARS), and a three-dimensional computational fluid dynamics (CFD) code (FLUENT). A visualization experiment has also been performed using a scaled-down model of the APR1400. The MARS analysis has predicted a serious core bypass phenomenon of borated water, while the CFD analysis has shown results opposite to the MARS results. The CFD analysis has shown that the flow pattern in the downcomer is fully three-dimensional and that vortex flow structures are formed near the cold legs so that the borated ...

International Nuclear Information System (INIS)

This paper consists of shielding analysis of steel balls and water filled end shields of Indian Pressurized Heavy Water Reactors (PHWRs). The material composition inside lattice tube is entirely different neutronically as compared with the composition of end-shield. Due to variation of material composition in radial and axial directions and complex geometry, it is necessary to carry out 3-D analysis for reasonable prediction of neutron flux and gamma dose rates. In the present paper, shielding analysis of end-shield for 540 MWe PHWR has been carried out during reactor operating and shutdown conditions using Monte-Carlo code MCNP. Furthermore materials on the periphery and central portion of end shield are different. Therefore the analysis was carried out separately for annular portion and central portion of end shield. ...

Energy Technology Data Exchange (ETDEWEB)

In the study a preliminary mid-term analysis of the BARCOM test model is presented. The BARCOM test model is a 1:4 scale of an existing pressurized heavy water reactor (PHWR) pre-stressed concrete inner containment of 540 MW Tarapur Atomic Power Station 3 and 4 units in India. The goal of this midterm analysis is to illustrate the modeling approach and achieve a prediction of the failure mode. The analysis was carried out using ABAQUS/CAE and ABAQUS/EXPLICIT version 6.7-EF1 software

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)

... Development Center, National Tsing-Hua University, Hsinchu, TW (China)

Energy Technology Data Exchange (ETDEWEB)

Energy savings in the synthetic fiber industry could be realized by using autoclave reactor condensate and boiler flue gas heat recovery. The non-cellulose (polyester) production process analysis shows that condensate returning from the reactor to the steam boiler raises inlet temperature, giving a reduced fuel requirement of about 8%. Also, boiler flue gas with a sufficiently high outlet temperature for boiler feed water and combustion air preheating results in further fuel savings. The process with an economizer saves up to 8.44%, and with a combustion air preheater, 6.25%. (Author)

International Nuclear Information System (INIS)

... 1343-3563 v. 57(1) SPECIFIC NUCLEAR REACTORS AND ASSOCIATED

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)

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

International Nuclear Information System (INIS)

Within the APROS simulation environment we have built a thermo-hydraulic model of a conceptual fusion power plant which is water cooled and uses lithium-lead for tritium breeding. For the safety assessment of this design we have studied an accident sequence which starts from a loss or coolant flow then leads to first wall breach and pressurisation of the vacuum vessel. Simulations have revealed strong pressure transients which can be alleviated by design changes. One goal is to verify the adequacy of the containment design: it remains intact at least 14 h without any mitigating efforts. Estimates for radioactive releases are obtained. (author)

Energy Technology Data Exchange (ETDEWEB)

Scenarios are developed for spent fuel that is taken out of the nuclear reactor and stored for 20 years before reprocessing, and three storage techniques which use a water pool, dry cask or vault are compared with respect to their costs. The storage price (storage cost per kilogram of spent fuel), which is employed as the economic index, is calculated on the assumption that all the charge is paid when the spent fuel is brought in the storage facilities. Four scenarios are assumed for spent PWR and BWR fuels to be stored in at-reactor (AT) or away-from-reactor (AFR) facilities. The capital costs cover the buildings of the storage facilities, equipment, decommissioning, casks for transportation and storage (for cask storage) and casks for transportation (for water pool or vault storage). Costs for operation and maintenance of the facilities are also considered in evaluating these ...

Energy Technology Data Exchange (ETDEWEB)

In this paper the diffusion of tritiated water vapor into concrete walls is studied to evaluate tritiated water retention capacity of a fusion reactor concrete building. Using a model of the tritiated water diffusion determined form experimental results, depth profiles of tritiated water in concrete are calculated in the case of being exposed to air containing tritiated water vapor during the normal operational condition of a fusion reactor. A 0.5-m-thick concrete is sufficient for reactor hall walls from a viewpoint of the tritium containment.

Energy Technology Data Exchange (ETDEWEB)

The development of KNGR (Korean Next Generation Reactor) is now in progress. KAERI is developing KNGR which is a advanced active PWR (pressurized water reactor) and 1350 MW electric capacities and is by based on UCN(Ulchin) 3 and 4 nuclear power plant which is a Korean standard PWR. In this report, the PZR (pressurizer) and Related System Design Features for KNGR which include PZR volume, PPCS (pressurizer safety valve)were analyzed. First, the Design Parameters between KNGR compared to UCH 3 and 4 were compared, and second, advanced design features of KNGR compared to UCN 3 and 4 were analyzed. After the present analysis, it has been concluded that the safety margins for the PZR level and pressure of KNGR were more increased by the larger PZR volume than those of UCN 3 and 4, for PZR minimum water level at reactor/turbine trip and PZR maximum pressure at ...

Science.gov (United States)

Computer programs have been developed to define the temperature increase which would be needed to bring deep-ocean water into density equilibrium with surface water for locations where data are available. A series of continuous-flow studies on phytoplankton blooms resulting from mixtures of 80 percent deep and 20 percent surface water in 2000-liter concrete culturing vessels (''reactors'') has been completed. A quantitative determination of nutrient utilization and flow through a combined primary and secondary trophic level system has been completed. This study utilized the clam Tapes semidecussata, fed from phytoplankton grown in 80 percent deep and 20 percent surface water. An analysis of the fate of the deep water discharged from a floating OTEC plant indicates that horizontal containment of the ...

Energy Technology Data Exchange (ETDEWEB)

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used ...

Energy Technology Data Exchange (ETDEWEB)

Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm{sup 3} H{sub 2}O/cm{sup 2}-s and decontamination ...

International Nuclear Information System (INIS)

BARC has organized an international round robin analysis program to carry out the ultimate load capacity assessment of BARC containment (BARCOM) test model. The test model located in BARC facilities Tarapur, is a 1:4 scale representation of 540 MWe pressurized heavy water reactor (PHWR) pre-stressed concrete inner containment structure of Tarapur Atomic Power Station (TAPS) unit 3 and 4. The features of the BARCOM test model and the constitutive data for the pre-test analysis along with the comparison of the results submitted by various participants are described in this pre-test report of the round robin analysis of BARCOM test model

International Nuclear Information System (INIS)

Purpose: To stably control the reactor water level so as not to cause excess water feeding in a BWR type reactor. Constitution: A flow control valve is disposed to the exit of a feedwater pump for a nuclear reactor and the valve is controlled by a flow regulator to maintain the water level constant in the reactor. A signal from a water level controller is inputted to the flow regulator to thereby control the flow rate control valve. In this case, the flow regulator remains in a saturated state just after the starting of the feedwater pump, in which the pump flowrate is at 100% to result in an excess water feeding condition. In view of the above, a feedback circuit is provided to the flow regulator so that the saturated state is eliminated and the water feeding can be controlled ...

International Nuclear Information System (INIS)

The device of the present invention improves reactor safety by suppressing lowering of water level in a shroud which surrounds a reactor core, even upon occurrence of rupture of pipelines in an emergency reactor core cooling system in a recycling pump-incorporated type reactor. Namely, an opening of each of cooling systems which forms the emergency reactor core cooling device in a reactor pressure vessel is disposed above the upper end of the reactor core. Further, it also comprises an independent high pressure water injection system, gravitational dropping type water injection system and an automatic depressurization system. With such a constitution, even if rupture of pipelines in the system should be assumed, coolants never flow directly from the shroud which surrounds the ...

International Nuclear Information System (INIS)

THe high-temperature reactor is one of the alternatives for the now predominantly employed water-reactors. In a recently published book designer Rudolf Schulten outlines his concept. In this article the book is reviewed. (author). 1 ref.; 1 fig.

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 ...

International Nuclear Information System (INIS)

Cooling water is sprayed on the outer surface of an upper portion of a container, and a pool is formed by the cooling water flowing down while cooling the container. Further, the cooling water stored in the cooling water pool is recycled by a pump for spraying the cooling water on the outer surface of the upper portion of the container. Sufficient amount of cooling water is supplied for spraying the cooling water to the outer surface of the upper portion of the container so that the outer surface of the container is free from drying and a liquid membrane is formed on the entire surface. The amount of the cooling water is made greater than that of the cooling water evaporated when the entire amount of the heat generate in the reactor core of the reactor is ...

International Nuclear Information System (INIS)

Actinides other than the main uranium or plutonium isotopes take a growing part in the different stages of the nuclear cycle. For the French nuclear power program based on the development of light water reactors and fast breeders, many evaluations of the secondary actinides build up are made for the both reactor types using mainly the existing reactor codes. The comparison of these foreseen compositions with experimental results allows to perform some adjustments of the neutronic data. The secondary actinide compositions are given for some typical fuels and their consequences on the nuclear cycle are discussed. An hypothetical burning of these wastes in fast reactors has been studied and the main conclusions are reported.

International Nuclear Information System (INIS)

... 2432 v. 43(1) SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS

Energy Technology Data Exchange (ETDEWEB)

The marine reactor MRX, an integral typed PWR with 100 MWt adopts one-through steam generators with coiling tubes. The cold feed water enters the steam generator and the super heated steam flows out. To avoid occurrence of flow instability in the steam generator due to a density wave oscillation, it is necessary to increase of flow resistance at the feed water inlet. The magnitude of flow resistance to stabilize the flow is determined by a simple linear analysis using a D-division method, of which accuracy is clarified by comparison with SRI's experiment. The external force due to heaving, one of ship motions will affect the flow behavior. Analysis by a modified RELAP5 capable of simulating the ship motions reveals that the effect of heaving becomes especially greater when the state of flow approaches both the conditions of density wave oscillation occurrence and resonance ...

Energy Technology Data Exchange (ETDEWEB)

This article gives an account of the recent development of light water reactors new concepts in the world. Different projects are being studied. The CE80+ from Combustion Engineering (CE) is a 1350 MWe-PWR-type reactor whose primary circuit is confined in a spherical metallic containment. This reactor was certified by NRC (national regulatory commission) in mid-1996. The APWR (advanced pressurized water reactor) is developed by MHI (Mitsubishi heavy industries) in a collaboration with Westinghouse, this PWR-type reactor fitted with 4 loops derived from the SP90 model that was developed by Westinghouse during the eighties. 2 units of ABWR (advanced boiling water reactor) were commissioned in Japan in 1996 and 1997, ABWR was certified by NRC in mid-1996. The BWR90+ is developed by ABB-atom (Sweden) and ...

International Nuclear Information System (INIS)

Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the ...

International Nuclear Information System (INIS)

A safety analysis code to design a pressurized water reactor and to obtain the licenses including entire proprietary rights is under development in domestic R and D project. The tasks of KAERI is to develop the constitutive relations including models for defining flow regimes and flow regime related models for inter-phase friction, wall frictions, wall heat transfer, and interphase heat and mass transfer in the two-phase three-field equations. In this paper, the process will be presented for choosing the best flow regime maps which occur in gas-liquid two-phase flow in horizontal and vertical tubes.

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code, named as SPACE, for a pressurized water reactor is under development to obtain a licensing to be used for the PWR design and to hold entire proprietary rights. The task of KAERI is to develop the physical models and correlations which are required to solve the field equations. It can be divided into four parts; i) flow regime determination, ii) wall heat transfer, iii) wall and interfacial friction, iv) interfacial heat and mass transfer. This paper will describe the process to develop the models for the two-phase flow regime maps in the horizontal and vertical pipes.

Energy Technology Data Exchange (ETDEWEB)

The MEX-15 is a conceptual design of a Multipurpose Reactor with thermal power of 15 MW and this reactor is pool type with fuel plates U{sub 3}0{sub 8}-Al of low enrichment uranium. This report presents the static calculation for the MEX-15 reactor using SRAC code system and was developed under the collaboration agreement between ININ-JAERI in Research Reactor Technology Development Division of Department of Research Reactor in Tokai Research Establishment. (Author)

Energy Technology Data Exchange (ETDEWEB)

The steam generator low water level trip setpoint of KORI NPP units 3 and 4(KNU 3 and 4), three-loop Westinghouse pressurized water reactor, is higher than that of OPR1000. In addition, steam generator downcomer water level in KNU 3 and 4 could fluctuate easily during a transient because of smaller downcomer water inventory, compared to the total water inventory in the steam generator. Due to these reasons, there is a higher possibility of unnecessary reactor trips caused by the steam generator low-low water level in KNU 3 and 4. Its operating history shows that most of reactor trips were caused by steam generator low-low level reactor trip signal. Such reactor trips, especially unnecessary ones, result in time and economic losses. In this paper, a ...

International Nuclear Information System (INIS)

This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume II: Monte Carlo Calculations for Nuclear Reactors. In-Core Management of Four Reactor Types. In-Core Management in CANDU-PHW Reactors. Reactor Dynamics. The Theory of Neutron Leakage in Reactor Lattices. Index.

International Nuclear Information System (INIS)

One of inherent characteristics of CANDU reactor is positive void reactivity in contrast to other pressurized light water reactors. During the large break loss of coolant accident, power pulse will be occurred during short time of early phase of accident due to positive void reactivity. However the duration of this power pulse is short, energy due to power pulse would be accumulated in the cladding material and will affect the peak cladding temperature or number of failed fuel elements. Recently, Canadian Nuclear Safety Commission (CNSC) indicated that the amount of void reactivity might be larger than the assumed values in safety analysis and this indication was based on the experimental data from ZED-2 facility. Based on that, the estimation of uncertainties due to the void reactivity during LBLOCA is the most important issue for CANDU safety analysis. In this study, a framework ...

International Nuclear Information System (INIS)

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for evaluation of debris in-vessel retention. The debris coolability analysis module is developed for the accurate prediction of the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with a temperature distribution and failure evaluation of the vessel wall. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis. The calculation for spreading is compared with a water spreading experiment on the floor and the calculation for coolability is compared with a n-octadecane melting experiment in a rectangular vessel. The comparisons show capabilities for predictions of ...

Energy Technology Data Exchange (ETDEWEB)

In this paper the possibility of configuring a water cooled Nuclear Thermal Propulsion (NTP) rocket, based on a Particle Bed Reactor (PBR) is investigated. This rocket will be used to operate on water obtained from near earth objects. The conclusions reached in this paper indicate that it is possible to configure a PBR based NTP rocket to operate on water and meet the mission requirements envisioned for it. No insurmountable technology issues have been identified.

International Nuclear Information System (INIS)

In a 540 MWe PHWR reactor at TAPP-4 the pressuriser has been incorporated in the PHT pressure control system to provide the necessary vapour cushion for PHT main circuit to reduce pressure variations due to transients involving swell and shrinkage. Need for the Pressuriser is due to the large inventory in PHT main circuit and relatively large heat source. The incorporation of Pressuriser is one of the modifications while upgrading from prevalent operating 220 MWe reactors to the present 540 MWe at TAPP-4. The sizing, design and selection of the 540 MWe Pressuriser has been done w.r.t swell/shrinkage requirement during normal/transient operation of the reactor. In this paper the commissioning and operating experience of the Pressuriser in conjunction with the PHT pressure control system is presented. During light water commissioning of PHT circuit some major hurdles like failure of pressuriser heaters ...

Science.gov (United States)

An advanced experimental technique has been developed to simulate two-phase flow behavior in a light water reactor (LWR). The technique applies three kinds of methods; (1) use of sulfur-hexafluoride (SF6) gas and ethanol (C2H5OH) liquid at atmospheric temperature and a pressure less than 1.0MPa, where the fluid properties are similar to steam-water ones in the LWR, (2) generation of bubble with a sintering tube, which simulates bubble generation on heated surface in the LWR, (3) measurement of detailed bubble distribution data with a bi-optical probe (BOP), (4) and measurement of liquid velocities with the tracer liquid. This experimental technique provides easy visualization of flows by using a large scale experimental apparatus, which gives three-dimensional flows, and measurement of detailed spatial distributions of two-phase flow. With this technique, we have carried out experiments simulating two-phase flow behavior ...

International Nuclear Information System (INIS)

The LWR fuel performance analysis computer code, FRAPCON-1, are evaluated to investigate the performance of CANDU fuel elements loaded in Wolsung-1 reactor. The FRAPCON-1 models of neutron flux depression in fuel and of fuel-to-cladding heat transfer are modified, and the validity of fission gas release model for CANDU fuel is evaluated. And the heavy water properties are provided in calculating the heat transfer coefficient between cladding and coolant. By using the modified code, FRAPCON-1-CSK, the sensitivity studies are carried out for Wolsung-1 fuel element design parameters. The performance analysis is also performed for Wolsung-1 fuel elements. The calculated results are discussed in terms of LWR fuel design criteria because of unavailability of CANDU fuel design criteria. (Author).

Energy Technology Data Exchange (ETDEWEB)

A core-wide in-phase neutron flux oscillation, which took place, for example, at LaSalle-2 in the USA in 1988, is one of the nuclear-coupled thermal hydraulic instabilities in boiling water reactors (BWRs). In this study, an analysis has been performed focusing on the excitation of this type of instability in BWRs due to seismically induced resonance, within the scope of a point kinetics model. For this purpose, the TRAC-BF1 code has been modified to take into account the external acceleration in addition to gravity. As a result of this analysis, it is shown that reactivity insertion can occur accompanied by in-surge of the coolant into the core resulting from excitation. It is also shown that the amount of reactivity inserted largely depends on the degree of stability of the initial state and the amplitude of the seismic wave, whose frequency is the same as the characteristic frequency of the ...

International Nuclear Information System (INIS)

PANDA is a large scale facility for investigating the long-term decay heat removal from the containment of a next generation of 'passive' Advanced Light Water Reactors (ALWR). PANDA was used to examine the long-term LOCA response of the Passive Containment Cooling System (PCCS) for the General Electric (GE) Simplified Boiling Water Reactor (SBWR). The first PANDA test series had the dual objectives of demonstrating the performance of the SBWR PCCS and extending the data base available for containment analysis code qualification. The test objectives also include the study of the effects of mixing and stratification of steam and noncondensible gases in the drywell (DW) and in the suppression chamber or wetwell (WW). Ten tests were conducted in the course of the PANDA SBWR Program. The tests demonstrated a favorable and robust overall PCCS performance under different conditions. The ...

International Nuclear Information System (INIS)

Since 1984 the thermal-hydraulic code ATHLET has been also applied for the analyses of LOCA and transients in VVER plants. The specific design of these plants especially of the steam generator design requires a specific modelling of the phenomena which may occur under LOCA and transient conditions in these plants. Differences in design compared to the design of western reactors have been briefly listed. Specific phenomena occurring under small leak accidents are shortly described. The consideration of the simulation of the boiler-condenser mode illustrates the modelling requirements for a code which may be applied to the prediction of such a thermal-hydraulic behaviour. Facing the lack of experimental data, the reliability of the simulation has been discussed by means of plausibility studies based on the momentum balance for steam and water. In summary: The VVER reactors differ in design compared to ...

Energy Technology Data Exchange (ETDEWEB)

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are unusually rapid.

International Nuclear Information System (INIS)

Active Process Water (APW) system is provided as a unitized system in TAPP-3 and 4. Maintenance on APW system requires shutdown of this system. As shut down heat exchangers are fed by APW system; during APW system shutdown cold shutdown state cannot be maintained. Therefore safety analysis is done to optimize the duration of reactor shutdown (which means low decay heat) after which APW shutdown can be taken with minimum water supply to the shutdown heat exchangers. Based on this analysis, it is proposed in technical specification that APW system shutdown can be taken after 7 days of reactor shutdown with shutdown heat exchangers supplied with about 20 % of normal APW flow. With this configuration, PHTS, moderator, end shield, calandria vault water temperature can be maintained within limits. A design provision is made at TAPP-3 and 4 to ...

Energy Technology Data Exchange (ETDEWEB)

Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering ...

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).

Science.gov (United States)

... in the design of such devices as fusion reactors, magnetohydrodynamic generators, magnetically levitated vehicles, magnetic forming devices, and ...

Science.gov (United States)

(accumulates & stores brine for disposal). Separator. (separates water from purge gases). ? Purge pump periodically vent ... Reactor Health. Sensor. ( verifies reactor is operating w/n limits) ... Waste and Hygiene Compartment ...

International Nuclear Information System (INIS)

One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations ...

International Nuclear Information System (INIS)

Since the mid-1990s, nuclear power plants in Korea have experienced wall thinning, leaks, and ruptures of secondary side piping caused by flow-accelerated corrosion (FAC). The pipe failures have increased as operating time progresses. In order to prevent the FAC-induced pipe failures and to develop an effective FAC management strategy, KEPRI and KOPEC have conducted a study for developing systematic FAC management technology for secondary side piping of all Korean nuclear power plants. As a part of the study, FAC analyses were performed using the CHECWORKS code. The analysis results were used to select components for inspection and to determine inspection intervals on each nuclear power plant. This paper describes the introduction of the FAC analysis method and the wall thinning trend analysis results by reactor type, system, and water treatment amine. This paper also represents the ...

International Nuclear Information System (INIS)

The super simulator 'SAMPSON' has been developed to show that there exist certain safety margins for light water reactors under hypothetical severe accidents and to investigate realistic measures of accident management by simulating accidents with a parallel computer. Heat-up of fuel rods and release of fission products from fuels are important factors to evaluate source terms. Models for fuel rod heat-up, hydrogen production due to cladding oxidation and cladding deformation and failure in the core region have been developed in the fuel rod heat-up analysis module. Fuel temperatures were calculated by solving the heat conduction equation. The calculated results for fuel temperature and hydrogen production were compared with CORA-13 experiment results. The comparisons showed prediction capability for the heat-up of fuel rods. The fission product release analysis module incorporates with models for ...

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).

Energy Technology Data Exchange (ETDEWEB)

The {sup 252}Cf-source-driven noise analysis method has been used in measurements for subcritical configurations of fissile systems for a variety of applications. Measurements of 25 fissile systems have been performed with a wide variety of materials and configurations. This method has been applied to measurements for (1) initial fuel loading of reactors, (2) quality assurance of reactor fuel elements, (3) fuel preparation facilities, (4) fuel processing facilities, (5) fuel storage facilities, (6) zero-power testing of reactors, and (7) verification of calculational methods for assemblies with the neutron k < l. These previous measurements, performed with a wide variety of multiplying systems, demonstrated the usefulness of the method. The high sensitivity of noise-measured parameters to small changes in fissile systems has been observed in several measurements. This high sensitivity has been ...

KoreaScience (Korea)

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KoreaScience (Korea)

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Science.gov (United States)

developing a consensus PRA standard for non- associated guidance light water reactor applications that will address some aspects of PRA in design. ...

International Nuclear Information System (INIS)

More than twenty years after commencing commercial operation in 1983, Wolsong Unit 1(W1- NPP), the first CANDU Pressurized Heavy Water Reactor (PHWR) in Korea, has been undergoing refurbishment. Safety analyses were required to evaluate the safety of W1-NPP because significant amount of equipment has been refurbished. To evaluate the effectiveness of W1-NPP after these upgrades, new safety analyses were performed using the same technical standards of Wolsong Units 2, 3, 4 (W234-NPP) for Design Basis Accidents (DBA). The refurbished W1- NPP is expected to be licensed for full power operation based on the verified safety analysis results that are obtained by using the upgraded computer codes and newly adopted technical standards of W234-NPP

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.

International Nuclear Information System (INIS)

The SCDAP computer code, which is being developed under the sponsorship of the United States Nuclear Regulatory Commission, models the progression of light water reactor core damage including core heatup, core disruption and debris formation, debris heatup, and debris melting. SCDAP is being used to help identify and understand the phenomena that control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and interpretation of severe fuel damage experiments and data. Comparisons between SCDAP calculations and the experimental data showed good agreement. Calculated and measured bundle temperatures for SFD-ST were within 200 K for the entire bundle and within 20 K for maximum cladding temperatures. For ESSI-2, calculated and measured maximum cladding temperatures were within 50 K, and the extensive liquefaction and relocation that was calculated was in ...

International Nuclear Information System (INIS)

Site studies for CANDU type NPP began in a careful manner since 1982 as a first part of the Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. A team was charged to develop all packages of the necessary main studies. The first Romanian NPP CANDU 6 type reactor gone to erection on Cernavoda site, planned to have 5 units and, like Wolsong NPP, applied the same design for the nuclear island. For the BOP parts the ANSALDO-GE project was applied with a thorough concern about requirements raised by connection to NSP. The first mission of design and research multi-branch team was to adapt the NPP Cernavoda project having an open water cooling circuit 'once-through' to the new parameters of a close recirculation water cooling circuit. Also, the structural design was re-evaluated for the case of soft foundation strata instead of hard rock ones. The close recirculation ...

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 out the salient ...

Energy Technology Data Exchange (ETDEWEB)

This report details the status of the Biological Conversion of Synthesis Gas Project. The following tasks are described as being completed: (1) the test plan, (2) culture development, and (3) the mass transfer/kinetic studies. The bioreactor studies (Task 4) are underway. The continuous stirred tank reactor system for the conversion of H{sub 2}S to elemental sulfur using Chlorobium thiosulfatophilum has been studied for varying light intensities. The system was also modified to include both sulfur recovery and cell recycle using ceramic membranes. Studies were also performed to observe the effects of cell recycle using a polysulfone hollow filter membrane module. Work on Task 5, limiting conditions/scale-up, includes a scale-up study with three different size reactors to establish the optimum operating conditions for hydrogen production from synthesis gas by the biological water-gas shift reaction using the photosynthetic ...

International Nuclear Information System (INIS)

The release rate of a nuclide from a reactor or a radioactive waste disposal plant at the accident is not steady, but varies with time. The various parameters of a nuclide migration into environment vary also day after day, or with the seasons. In such cases, dynamic behavior of the nuclide in the environment must be taken into consideration. It is difficult for a mathematical model to involve all of mechanisms for the nuclide migration. The environment for evaluation of doses are usually divided into some of compartments in which a nuclide concentration is uniform. Time variations of the nuclide concentration in the compartment are described in simultaneous differential equations. The nuclide concentration can be solved as a time function, and the radiation doses, therefore, can be estimated as a time function. Generic analysis code for dynamic compartment model (GACOM) is developed for the nuclide migration and the evaluation of doses in ...

International Nuclear Information System (INIS)

In this paper, we report on the analysis of reverse flow in inverted U-tubes of a steam generator under natural circulation condition. The mechanism of reverse flow in inverted U-tubes of the steam generator with natural circulation is graphically analyzed by using the full-range characteristic curve of parallel U-tubes. The mathematical model and numerical calculation method for analyzing the reverse flow in inverted U-tubes of the steam generator with natural circulation have been developed. The reverse flow in an inverted U-tube steam generator of a simulated pressurized water reactor with natural circulation in analyzed. Through the calculation, the mass flow rates of normal and reverse flows in individual U-tubes are obtained. The predicted sharp drop of the fluid temperature in the inlet plenum of the steam generator due to reverse flow agrees very well with the experimental data. This indicates that the developed ...

Energy Technology Data Exchange (ETDEWEB)

The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

Science.gov (United States)

Dr. Rudolf Schulten of West Germany and expert witnesses from national laboratories, utilities, and the nuclear industry testified on reactor safety issues as they relate to the Chernobyl accident and public concern that modern technology has not paid enough attention to public safety. Each of the witnesses contributed safety-related information based on what has been learned from the Soviet incident. Particular focus went to similarities and differences between the Chernobyl and US reactors in safety design and engineering and to the environmental effects of the accident. The N reactor near Richland and a commercial reactor at Fort St. Vrain, Colorado are the only two operating graphite reactors, but neither is a boiling water reactor.

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 figs., 6 tabs.

Energy Technology Data Exchange (ETDEWEB)

This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the ...

Energy Technology Data Exchange (ETDEWEB)

This is the sixth book on reactor materials published under sponsorship of the Naval Reactors Office of the United States Department of Energy, formerly the United States Atomic Energy Commission. This book presents a comprehensive compilation of the most significant properties of thorium dioxide, much like the book Uranium Dioxide: Properties and Nuclear Applications presented information on the fuel material used in the Shippingport Pressurized Water Reactor core.

International Nuclear Information System (INIS)

Most of all research reactors are immerged in the deep water pool to be a ultimate heat sink. At the neighbor of the reactor, some radio-active matters, such as Na-24, Ar-41, Mg-27, Al-28 and etc, may be generated by the neutron irradiation. Those radio-active isotopes may rise up to the pool water surface through the natural convection flow, which can make the radioactivity in the reactor hall rise high enough to concern about the health of people working in the reactor hall. When the irradiation test facilities are loaded or unloaded during a normal operation, the highly radio-activated primary coolant may flow out through the irradiation test holes on the top of the reactor. This also may be a main hazard source to make the working environment of the reactor hall bad. Making a hot water layer 1.5 ...

International Nuclear Information System (INIS)

The corrosion rate of low alloy steel SA-508 and carbon steel A-410b in simulated operation and shutdown conditions of pressurized water reactor has been determined Moreover potentiodynamic polarization curves and galvanic effect through coupling of AISI-304 have been carried out under shutdown simulated condition. (Author) 8 refs.

International Nuclear Information System (INIS)

Purpose: To eliminate the water level deviation due to the recycling flowrate, as well as enable a stable control to a reference value even upon changes in the recycling flowrate caused by the variation in the opening degree of a minimum flow valve. Constitution: Reactor recycling system comprises a feedwater pump, a flowrate control valve, a reactor water level detector, and a minimum flow line and a minimum flow valve for preventing the overheating of the feedwater pump at a low flowrate. A flowrate compensator is further disposed, in which a recycling flowrate signal is subtracted from a pump flow rate signal and the result is fedback as a compensated pump flowrate signal. This enables the control system to operate at a rapid response for suppressing the effect of the recycling flowrate as external disturbance, whereby the water level in the reactor can be ...

International Nuclear Information System (INIS)

During the last years one focus of German PWR safety analysis was boron dilution events with the potential of reactivity transients. Coolant with a low boron concentration could be collected in localized areas of the reactor coolant system e.g. by separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux- condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. The TUeV NORD SysTec was charged by German supervisory authorities with the assessment of the safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses shall demonstrate that boron dilution events cannot lead to recriticality of the core. Hence the ...

Energy Technology Data Exchange (ETDEWEB)

A highly reliable control rod drive mechanism (CRDM) installed inside the reactor vessel has developed for use of an advanced marine reactor. This CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. The CRDM works in the high temperature and high pressure water - 310degC and 12 MPa, the same atmosphere as the primary loop. Driving force is produced by a synchronous motor with the rotor of a permanent magnet, which has been developed. An innovative latch mechanism using separable ball nuts can latch driving shaft connecting the control rod and de-latch it for scram. The rod position detector using a magnetostrictive wire type sensor on the principle of Wiedeman effect has been developed, accuracy of which is verified to have a detecting error within 1.2 mm. Ball bearings for thrust and radial supports in rotation have been ...

Energy Technology Data Exchange (ETDEWEB)

The prediction of two-phase flow regime in the horizontal pipings during a loss-of-coolant accident (LOCA) is important for safety analysis of a pressurized water reactor (PWR). The flow regime transition conditions for a horizontal two-phase air-water flow were studied using a large-height, horizontal rectangular duct test section. The duct dimensions were 700 mm in height, 100 mm in width and 28.3 m in length. The experimental criterion for the flow regime transition from the stratified to slug flow regimes, in terms of the local void fraction and the non-dimensional gas-liquid relative velocity, agreed qualitatively with the prediction by the Mishima-Ishii model that is based on an idea that the interfacial waves with the largest growth rate will develop into a slug. However, the transition in the experiment occurred at systematically lower (by about 40 %) relative velocities than the prediction by ...

Energy Technology Data Exchange (ETDEWEB)

This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the ...

KoreaScience (Korea)

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Energy Technology Data Exchange (ETDEWEB)

Real-time neutron radiography is being evaluated for studying the dynamic behavior of two phase flow and for measuring void fraction in vertical and inclined water ducts. This technique provides a unique means of visualizing the behavior of fluid flow inside thick metal enclosures. To simulate vapor conditions encountered in a fluid flow duct, an air-water flow system was constructed. Air was injected into the bottom of the duct at flow rates up to 0.47 I/s (1 cfm). The water flow rate was varied between 0--3.78 I/m (0--1 gpm). The experiments were performed at the Pennsylvania State University nuclear reactor facility using a real-time neutron radiography camera. With a thermal neutron flux on the order of 10{sup 6}n/cm{sup 2}/s directed through the thin duct dimension, the dynamic behavior of the air bubbles was clearly visible through 5 cm (2 in.) thick aluminum support plates placed on both sides of ...

International Nuclear Information System (INIS)

Active Process Water (APW) System, provided as unitized closed loop system in Tarapur Atomic Power Project Units-3 and 4, serves to remove heat from various heat exchangers. One of the important loads served by APW system is shutdown cooling heat exchangers and if APW shutdown is taken then reactor cannot be maintained in cold shutdown condition. It is estimated that after 7 days of reactor shutdown, if about 20% of the normal cooling flow to shutdown cooling heat exchangers is provided then along with keeping PHT in cold shutdown state, reactor components, moderator, end shield water, calandria vault water and calandria vault concrete temperature can be maintained within technical specification limits for extended duration. (author)

International Nuclear Information System (INIS)

Modelling the behaviour of fission product (FP) in a nuclear reactor coolant system (RCS) undergoing a hypothetical severe accident is an important step in the evaluation of radioactive release outside a nuclear power plant. This paper scrutinize Small Break LOCA sequence for WWER1000 reactor in order to investigate the possible paths for release of FP from fuel pallets to the reactor containment. Contemporaneous computer code for simulation of RCS will be use for the analysis. The results from analysis of fuel damage and release of FP trough the break of cold leg are present. (author)

British Library Electronic Table of Contents (United Kingdom)

In allusion to the explosion of a urea reactor took place in a fertilizer plant at Pingyin, Shandong, China, a series of evidence collection and inspection jobs which includes collecting operation condition and parameters, sampling the explosion fracture, reactor body apart from explosion fracture, and leak detection medium and its hangover, etc., had been carried out firstly. Based on these jobs, farther analysis and computation work has been done to the structural and materials characteristics and the operation condition of the urea reactor, including compositions, metallographic phases, tensile properties, impact energy, strain ageing characteristics, and fracture toughness of the urea reactor steels, the compositions of leak detection medium and its hangover in the urea reactor, and ex...

Energy Technology Data Exchange (ETDEWEB)

The calculated radioactive waste inventories of the Turkey Point pressurized water fission reactor (PWR) and the Starfire conceptual fusion tokamak are compared as a function of time from initial start-up to 10,000 years after decommissioning. Only material out of reactor at least one year is considered. The total activity in Ci/W(th) of the Starfire tokamak is slightly greater than that of the PWR during the active lifetimes of the two reactors and beyond 1000 years. However, using reduced activation materials in Starfire can result in about 1/2000 as much long-lived radioactivity as in the fission reactor. It is stressed that comparison of wastes on this basis is not straightforward, since the radioisotopes and methods required for their disposal are different for fusion and fission reactors. 2 refs., 1 fig., 2 tabs.

Energy Technology Data Exchange (ETDEWEB)

This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid ...

Energy Technology Data Exchange (ETDEWEB)

The five thermal-hydraulic concepts chosen for conceptual study of advanced PWR systems have been studied as follows: (1) Critical Heat Flux in passive PWR Conditions: review of previous works (various of correlations, analysis of parametric trends) on CHF, assessment and improvement of CHF prediction models for round tubes, development of the prediction model on bundle CHF with considering the correction factor calculated from the tube data base, design and construction of the intermediate-pressure CHF experimental loop, extension of CHF data base by performing the experiments at low-flow, and low-quality conditions (2) Passive Cooling Concepts for Concrete Containment Systems: Selection of the external condenser by comparing and reviewing between passive cooling concepts for concrete containment system concepts, survey and review of previous studies (theoretical mechanism of condensation heat transfer and effect of non-condensable gases) on the condensation ...

Energy Technology Data Exchange (ETDEWEB)

A model was developed to compute concentration of impurities in the WWER 440 steam generator (SG) secondary water along the tube bundle. Calculated values were verified by concentration values obtained from secondary water sample chemical analysis. (orig.). 2 refs.

International Nuclear Information System (INIS)

The present paper deals with the experimental determination of pressure drop across a four-cusped vertical channel. This geometry represents, ideally, the blockage condition in a typical pressurized water reactor with core degraded by accident. Experiments were performed for both single and two-phase flow. Water was utilized for the single-phase measurements whilst simultaneous flow of air and water simulated the steam-water flow. Observation of the prevailing two-phase flow regime was carried out, so that its mechanism could be fully understood. The averaged void fraction was also measured, by the gamma-ray attenuation technique. A wide range of water and air mass flow rates was covered, so that all flow conditions, possible to exist in a reactor with LOCA, could be investigated. New correlations for pressure drop are proposed. (Author).

International Nuclear Information System (INIS)

Radionuclide X-ray fluorescence analysis was used for the determination of iron, calcium, potassium, copper, manganese and zinc in waste water from the production process of citric acid. (author).

Science.gov (United States)

Mar 1, 2011 ... analysis of NaMBT in ethylene glycol-water solutions with high accuracy. It reduces analysis time, requires smaller samples, and is able to ...

Energy Technology Data Exchange (ETDEWEB)

This report describes a simple seismic analysis model of the KALIMER-600 sodium cooled fast reactor and its application to the seismic time history analysis. To develop the simple seismic analysis model, the detailed 3-D finite element analyses for main components, IHTS piping system, and reactor building were carried out to verify the dynamic characteristics of each part of simple seismic analysis models. By using the developed simple model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design of KALIMER-600 were performed. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity.

International Nuclear Information System (INIS)

... Toshiba Corporation, Isogo Engineering Center, Yokohama, Kanagawa (Japan)

KoreaScience (Korea)

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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)

Science.gov (United States)

An optimal deployment pattern of flux mapping detectors for a Canada uranium-deuterium (CANDU)-600 pressurized heavy water reactor (PHWR) is determined by obtaining an optimal feedback relationship between flux measurements and zone controllers. The reactor core is modeled with a time-dependent two-group, two-dimensional diffusion equation, and flux perturbation are expressed by model expansions. The modal expansion coefficients are used as elements of the state vector representing the system dynamics. An optimal feedback matrix connecting the flux measurement vector to the control vector is derived by minimizing a quadratic performance index involving both the state and control vectors. We obtain the detector effectiveness in terms of the optimal feedback matrix and determine optimal detector locations for the Wolsung Unit 1 reactor in Korea. We have tested the methodology through evaluation of flux ...

International Nuclear Information System (INIS)

In Pressurised heavy water reactors (PHWRs), neutron poison in the moderator is used to compensate the excess reactivity present in the core on different occasions such as xenon decay during synchronization just after poison out period or start ups from xenon free conditions. It is also used in secondary shutdown system (SDS-2), where required amount of neutron poison is injected directly into the moderator within 2.5 seconds. Further, it is also used for over poisoning the moderator to achieve the guaranteed shutdown state when the regular shutdown systems are taken for maintenance. Generally, two types of moderator poisons are used in power reactors to balance the reactivity of the core and they are boron and gadolinium. Gadolinium is used in the form of gadolinium nitrate (Gd(NO3)3.6H2O). The paper gives the details of estimation of reactivity coefficients of gadolinium for 540 MWe PHWR for different operating ...

International Nuclear Information System (INIS)

The emittance of aluminum components exposed to low-temperature aqueous solutions were required for thermal analysis of a Loss of Cooling Accident for the Savannah River Site production reactors. Experimental data for the thickness and emittance of oxide films formed under these conditions were collected and reviewed. Correlations were developed for the oxide film thickness and corresponding total hemispherical emittance. Film thickness and emittance were also measured for the specific conditions of interest in order to verify the predictions based on the literature data. After one year of exposure in 30deg C reactor moderator, the aluminum oxide film thickness is predicted to be 6.4 #mu#m#+-#10%; this value is relatively insensitive to exposure time. Some phenomena which would tend to yield thicker oxide films in the reactor environment relative to those obtained under experimental conditions were ...

Energy Technology Data Exchange (ETDEWEB)

In light of the recent rapid increase in the fossil fuel prices it is meaningful to evaluate the impact of these price changes in the economics of dual-purpose desalination projects producing electricity and fresh water simultaneously. The price of crude oil and LNG (Liquefied Natural Gas) has increased by about 200% and 100% during the past three or four years. The uranium price has also increased by nearly 500% during the same period. The purpose of this paper is to analyze and compare the economics of SMART (System-integrated Modular Advanced ReacTor) which is being developed as a small size PWR type and the LNG Combine Cycle coupled with MED (Multi-Effect Distillation) which are being acknowledged as promising energy sources for the future in Korea. The methods of analysis used in this paper are the lifetime leveled cost method for the power and water cost calculation and the power credit method for ...

International Nuclear Information System (INIS)

The High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory is one of the world's most powerful research reactors. In 1996, one year after the demise of the Advanced Neutron Source Project, the U.S. Department of Energy embarked on an aggressive program to upgrade the neutron scattering facilities at the HFIR. These upgrades, which are now in progress, include the installation of larger beam tubes, a high-performance hydrogen cold source, and additional neutron guides and neutron scattering instruments. An extensive analysis effort was performed over the past 4 yr to support the design of the modified beamlines and new user facilities and to assess the impact of the upgrades on the integrity of the existing reactor system. The results of three of these analyses are summarized here. Specifically, results are presented for analyses related to the design of the new cold neutron source ...

International Nuclear Information System (INIS)

Purpose: To remove excessive driving pressure applied to an unisolated control rod drive by returning excessive coolant to a condensed water storage tank or to the inlet side of a drive water pump using a coolant flow rate control pipe of a control rod driving hydraulic system. Constitution: Excessive water is returned to a condensed water tank while controlling the excessive coolant by a flow control valve in response to variations in the pressure difference between the reactor pressure and the driving water line when the control rods are isolated using a pipe from the outlet side of the drive water pump to the condensed water storage tank. Thus, the control rod to be isolated is prevented form being dropped. (Sekiya, K.).

International Nuclear Information System (INIS)

The Advanced MAPLE is a research reactor design under development as a high-flux neutron source. The main performance goals for the reactor are a high peak thermal neutron flux in a heavy-water reflector tank, and a high average fast neutron flux in a central irradiation facility, with a maximum linear fuel rod rating of less than 120 kW/m. This study investigated the neutronic and reactor design consequences of the use of H_2O coolant as opposed to D_2O. The neutronics results, and several other considerations, indicate that H_2O coolant has a number of advantages. It is suggested that the H_2O coolant option be considered in the design of the Advanced MAPLE reactor. (L.L.) 9 refs., 4 figs., tab.

International Nuclear Information System (INIS)

The risk analysis was carried out in consideration of conditions prevailing at the Kalkar site analogous to the investigations in phase A of DRS (German Reactor Study). Earthquake design loads include the probabilities of upper deviations of the site intensities to be expected. The calculations of dynamic loads for select buildings are made using models and computational methods. Component analyses were performed analogous to DRS for the supports of large components, supports of the roof construction of the reactor building taking into account support reserves due to plastic work capacity, wall disks in steam generator buildings and switchboard plant buildings. (DG).

Energy Technology Data Exchange (ETDEWEB)

In the present paper, experimental study of forced convection boiling were performed to investigate the post-CHF characteristics of a vertical annular channel with one heated rod and four spacer grids for new refrigerant R-134a. The experiments were conducted under outlet pressure of 11.6, 13, 16 and 20 bar, mass fluxes of 100-600 kg/m{sup 2}s, and inlet temperatures of 25-51 .deg. C. The parametric trend of the post-CHF data was well consistent with previous studies. The two phase flow regime in tube flow occurring downstream of the CHF has been called post-CHF, dispersed flow, liquid-deficient flow, mist flow and film boiling. This regime is characterized by a continuous vapor phase with discrete liquid drops and a non-wetted heated surface. This regime has a considerable importance in the areas of light water reactor(LWR) accident analysis and other film boiling applications. The post-CHF region occurs by design in heat ...

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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 the validity of the ...

International Nuclear Information System (INIS)

Eight years ago, corrosion and tube denting seriously threatened the reliability and design life of steam generators, especially for closed loop arrangements in pressurized water reactors (PWRs). Concentrated research by the Steam Generator Owners Group (SGOG) diagnosed the causes and produced effective solutions, notably guidelines for water chemistry control in the secondary loop. The guidelines recommend specific levels of water impurities and remedial actions to prevent cooling-water leaks in the condenser, prevent air leaks, limit corrosion product buildup, and remove some impurities while neutralizing others. Continued research in SGOB-II is investigating intergranular corrosion and stress corrosion cracking. 3 figures.

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An inorganic particulate adsorbent of a titania-alumina is described for treating a superheated water containing radioactive materials such as cobalt ions, which is free from release of corrosive impruities, and which has a high adsorption capacity of radioactive materials and a high mechanical strength is prepared by hydrolyzing a titanium alkoxide and an aluminum alkoxide, thereby forming a hydrous titanium oxide and a hydrous aluminum oxide, respectively; precalcining the hydrous titanium oxide and aluminum oxide, mixing and molding the resulting titania and alumina into a particulate mixture thereof having a titania mole fraction of 0.2 to 0.9, and calcining the particulate mixture at 500/sup 0/-700/sup 0/C. This absorbent is effectively used in treat boiling water-type nuclear reactor core-circulating water to remove radioactive substances therefrom.

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... 978-5-94883-072-8 121 p. SPECIFIC NUCLEAR REACTORS AND

Energy Technology Data Exchange (ETDEWEB)

As evidence of its effectiveness rapidly accumulates, the Lomi process has most recently been used to decontaminate the recirculation loops and the reactor water clean-up unit of a BWR at Monticello in the United States. An average decontamination factor of 23 was achieved in the recirculation loops.

Science.gov (United States)

While tritium exposure to the site-workers in Wolsung NPP is up to about 40% of the total personnel exposure, Ministry of Science and Technology has asked tritium removal facility for requirement of post heavy-water reactor construction. For the purpose o...

International Nuclear Information System (INIS)

The development and successful application of a unique leak sealant formulation comprised of a mixture of graded, hollow ceramic microspheres, surface oxidized aluminum powder and saturated gibbsite suspension is described. The project was undertaken to address the escalating leakage from up to 15 small weld defects in the water reflector vessel, an integral component of the NRU (National Research Universal) reactor calandria. The reflector surrounds the reactor core with a neutron reflecting blanket of light water. Injection of the sealant is typically done with the reactor shutdown and the water reflector system operating normally, but can also be performed with the reactor at full power. The procedure is simple and effective. Individual treatments of as little as 125 ml of sealant (10 ppm in the 12,500 L system) have yielded leak ...

Science.gov (United States)

The Wolsung-type nuclear fuel localization project carried out since 1981 finally reached to a full-fledged phase in 1987. We successfully produced and timely delivered a yearly demand of nuclear fuel for Wolsung unit 1. In this report we studied and summ...

International Nuclear Information System (INIS)

The major feed-water line break accident is re-analyzed, which is based on Guangdong Daya Bay nuclear power station final safety analysis report, to justify the impacts of the decreasing of auxiliary feed-water flow rate on the safety margin in Daya Bay. The results showed that the accident analysis can meet the demands of acceptance criteria with the auxiliary feed-water flowrate decreasing from 45 m"3/h to 41.8 m"3/h, and enough safety margin is still retained

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).

Energy Technology Data Exchange (ETDEWEB)

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)

The radiation dose and the atmospheric dispersion for Pressurized Heavy Water Reactors (PHWR) are based on the CAN/CSA N288.2-M91 standards: for Pressurized Water Reactor (PWR) on the NRC Regulatory Guide 1.145. There are some differences between in the methodologies used in the standards, including the atmospheric dispersion model, the release height, the temperature lapse rate, the cutoff condition. This paper reports on a comparison of standards for atmospheric dispersion models of PHWRs and PWRs in order to determine which one is the more conservative. The comparison between PHWR and PWR for atmospheric dispersion factors and radiation doses confirms that there are no big differences

International Nuclear Information System (INIS)

Traditionally, probabilistic risk analyses [PRA] of severe accidents in nuclear power plants have limited themselves to consideration of the set of initiating events occurring during full power operation. However, some analyses of accident initiators during low power, shutdown, and other modes of plant operation other than full power have been performed. These studies as well as the Chernobyl accident and recent operating experience at U.S. pressurized water reactors suggested that risks during low power and shutdown could be significant. As such, the analysis of the frequencies, consequences, and risks of these accidents was identified as one task in the Nuclear Regulatory Commission staff's study of the implications of the Chernobyl accident to U.S. commercial nuclear power plants. The surry PRA project is an ongoing high priority effort at BNL [Brookhaven National Laboratory] that is expected to continue for the next ...

International Nuclear Information System (INIS)

This paper gives an overview of a multiyear joint research program being conducted at the University of New Mexico (UNM) with support from Sandia National Laboratories and GA Technologies. This research focuses on heat removal and fluid dynamics in flow regimes characterized by low pressure and low Reynolds number. The program was motivated by a desire to characterize and analyze cooling in a broad class of TRIGA-type reactors under: (a) typical operating conditions, (b) anticipated, new operating regimes, and (c) postulated accident conditions. It has also provided experimental verification of analytical tools used in design analysis. The paper includes descriptions of the UNM thermal-hydraulics test facility and the experimental test sections. During the first two years experiments were conducted using single, electrically heated rod in water and air annuli. This configuration provides an observable and serviceable ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this research was to determine the extent of damage that occurs when two pipes experience an impact event due to one whipping against the other. The research was conducted through experimental and analytical approaches. The former required the development of a specialized impact machine that could accelerate a whipping pipe with sufficient energy to cause failure of a target pipe that was heated and pressurized to Pressurized Water Reactor (PWR) conditions. Damage was measured in terms of crushing, bending, and failure. The results of the tests permitted the correlation between pipes of a certain size and the damage they could cause when impacting with a certain amount of known energy. These results were used to evaluate the pipe whip criteria in the Standard Review Plan 3.6.2-4. It was established that the criteria conditions did not fully represent the results obtained experimentally. An analysis ...

Energy Technology Data Exchange (ETDEWEB)

Biocatalytic Desulfurization (BDS) represents an alternative approach to the reduction of sulfur in fossil fuels. The objective is to use bacteria to selectively remove sulfur from petroleum and middle distillate fractions, without the concomitant release of carbon. Recently, bacteria have been developed which have the ability to desulfurize dibenzothiophene (DBT) and other organosulfur molecules. These bacteria are being developed for use in a biocatalyst-based desulfurization process. Analysis of preliminary conceptual engineering designs has shown that this process has the potential to complement conventional technology as a method to temper the sulfur levels in crude oil, or remove the recalcitrant sulfur in middle distillates to achieve the deep desulfurization mandated by State and Federal regulations. This paper describes the results of initial feasibility studies, sensitivity analyses and conceptual design work. Feasibility studies with various crude oils ...

Energy Technology Data Exchange (ETDEWEB)

Aqueous mineral carbonation is a potentially attractive sequestration technology to reduce CO2 emissions. In this paper, the energy consumption and costs of this technology were assessed using either wollastonite (CaSiO3) or steel slag as feedstock. The major energy-consuming process steps were found to be the grinding of the feedstock and the compression of the CO2. Within ranges of experimentally investigated process conditions, optimum energetic CO2 sequestration efficiencies were 79 and 74% for wollastonite and steel slag, respectively. It was shown that the energetic performance for both feedstock might be improved up to >90% by e.g. further grinding of the feedstock and reducing the amount of process water applied. At energetically optimized process conditions, a preliminary cost estimate was made of 93 and 66 euro/ton CO2 avoided for wollastonite and steel slag, respectively (sequestration costs excluding possible capture). For wollastonite, major ...

International Nuclear Information System (INIS)

The present paper is concerned with development and application of a so-called Effective Convection Model (ECM), which aims to provide a detailed, mechanistic description of heat transfer processes in a BWR lower plenum. The ECM is a Computational Fluid Dynamics (CFD)-like tool which employs a simpler and more effective approach to compute heat transfer by solving only energy conservation equation instead of solving the full set of Navier-Stokes and energy equations by a CFD code. We implement the ECM in a CFD code (Fluent), with detailed description of the ECM development, implementation and validation. A dual approach is used to validate the ECM, namely validation against experimental data and against heat transfer results obtained by CFD predictions in the same geometries and conditions. Insights gained from CFD simulations are also used to improve ECM. The ECM capability as an effective tool to simulate heat transfer of an internally heated volume in 3-dimensional complex geometry ...

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).

International Nuclear Information System (INIS)

Full text of publication follows: The heat transfer and flow in narrow channels has lots of advantages such as compact structure, high efficiency, design flexibility and so on. So it is widely used in the fields such as the new reactor core plate elements, the once-through stream generator, compact heat exchangers as well as electronic components. In recent years, more strong attentions have been attracted to the thermal-hydraulic characteristics and mechanism of the two-phase flow in narrow channels. As the flow regime characteristics of two-phase flow is fundamental one of them, the research on the two-phase flow regimes and the regime transitions in horizontal rectangular narrow heated channels can provide theoretical foundation and engineering directions to the whole research on the thermal-hydraulic characteristics and mechanism of the two-phase flow in narrow channels. The characteristics of two-phase flow regimes and regime transitions of boiling ...

Energy Technology Data Exchange (ETDEWEB)

The Korean Next Generation Reactor (KNGR) adopts 4-train Direct Vessel Injection (DVI) configuration and injects the safety injection water directly into the downcomer through the 8.5'' DVI nozzle. Thus, the thermal hydraulic phenomena such as ECCS mixing and bypass are expected to be different from those observed in the cold leg injection. In order to investigate the realistic injection phenomena and modify the analysis code developed in the basis of cold leg injection, thermal hydraulic test with the performance evaluation is required. Preliminarily, the sequence of events and major thermal hydraulic phenomena during the small break LOCA for KNGR are identified from the analysis results calculated by the CEFLASH-4AS/REM. It is shown from the analysis results that the major transient behaviors including the core mixture level are largely affected by the downcomer ...

International Nuclear Information System (INIS)

The research committee of the Atomic Energy Society of Japan on water chemistry standard aims at establishing the private standard of water chemistry of nuclear power plants. The committee gathers up 'BWR water chemistry management manual', 'PWR primary system water chemistry management manual' and 'PWR water chemical analysis standard method', and furthermore aims at the standardization of those in future. Looking back on the committee's activities for the past four years, latest results of research of water chemistry mainly contributing to safe and reliable nuclear power plants were described with the future perspective of water chemistry and a demanded break-through. (T.T.)

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 ...

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Ultratrace analysis of mercury in rain water has been investigated, including recovery of ionic mercury with the addition of acids, treatment and storage and stability studies of mercury in rain water. Recovery of ionic mercury from rain water samples increased with increased acid concentrations and increased substantially with UV-irradiation. Recovery of ionic mercury in the presence of different acids was compared. Treatment studies of rain water for the analysis of total mercury were done using different decomposition methods. More than 20% of the total mercury was found in rain water residue after filtration through 0.45 ..mu..m membrane filters. Methylmercury (MM) content in filtered and unfiltered rain water however was nearly the same. On long term storage of rain water without acids at pH = ...

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)

At present, two types of reactors, Pressurized Light Water Reactor(PLWR) and Pressurized Heavy Water Reactor(PHWR), are operating and under construction in Korea. They are much different in design concepts and inherent features from each other so that the calculation methods for Exclusion Area Boundary(EAB) are also different from each other. Thus, the domestic calculation methodology has been applied to PHWR, Wolsung 2, 3 and 4. In this report, the regulatory requirements and methodologies for EAB of Canadian methodology for EAB has been also investigated. It has been examined that the Canadian methodology which has been applied to the calculation of EAB of Wolsung 2, 3 and 4 can be said to be conservative enough compared to physical phenomena. 4 tabs., 3 figs., 22 refs. (Author).

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A new kind of receiver-reactor for high-temperature solar furnaces is proposed. The main body of the receiver component is an ellipsoid of revolution with specularly reflecting inner walls. The reactor component, a crucible, is placed at one focal point and the aperture at the other. With this arrangement, substantially all of the incident radiation from the concentrator should reach the reactor directly or after one reflection from the cavity walls. An analysis of the radiative exchange among the surfaces is presented. The analysis provides a tool for a parametric study and optimization of the design. It is found that, in contrast to that of conventional well-insulated cavity receivers, its collection efficiency is not very sensitive to the size of its aperture.

International Nuclear Information System (INIS)

Human errors have been reported as one of the most significant causes of major events in nuclear power plants (NPPs). For example, Kim and Park found that about 23% of the major events that occurred at NPPs in Republic of Korea from 1986 to 2006 were caused by human errors. For this reason, a detailed analysis on human errors is an important task for increasing the safety of NPPs. Kim and Choi?2 analyzed 100 human-related unplanned reactor trip events in the Republic of Korea from 1986 to 2006 to consider the type of human errors based on the simple path model for human-induced unplanned reactor trips developed by Kim and Park. In this paper, we will investigate and perform a detailed analysis of the data to identify human-related unplanned reactor trip trends

International Nuclear Information System (INIS)

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ''significant to safety,'' and identification of key operator actions for the analyzed events.

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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.

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 advanced nuclear power ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

The authors have developed a complex of activation methods of analysis using a nuclear reactor (nuclear activation analysis) and a cyclotron (charged-particle activation analysis). The methods have been used to determine the concentrations of more than 20 elements in five medicinal plants native to Uzbekistan: Syrian rue (Peganum harmala L.), plantain (Plantago lanceolata), peppermint (Mentha piperata L.), sage (Salvia officinalis L.), and ziziphora (Ziziphora bungeana Yur.). The results of radio-activation analysis were compared with the results of standard spectral analysis performed in another laboratory and the accuracy of the procedures developed was evaluated on the basis of the results.

International Nuclear Information System (INIS)

Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of ...

Energy Technology Data Exchange (ETDEWEB)

The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary instrumentation effort in support of Arms Control ...

International Nuclear Information System (INIS)

The Gross Decontamination Experiment was conducted on various levels and surfaces of the TMI-2 Reactor Building during February and March 1982 and was designed to investigate the effectiveness of various surface decontamination techniques. The polar crane, D-rings, missile shields, refueling canal, fueling bridge, major equipment, floors and some walls were flushed with low pressure water. Water lances were directed manually and applied water at temperatures between ambient and 60"0C at a flow rate of about 95 liters per minute. In addition, floor surfaces on the 305-ft elevation and floor surfaces and major equipment on the 347-ft elevation were sprayed with high pressure water (floors in the Reactor Building are designated by their elevations above sea level). The water pressure in this case varied between 13.8 and 41.4 mPa and ...

International Nuclear Information System (INIS)

Inherent boron dilution can occur in case of a Small Break LOCA when low borated water is mainly accumulated in the U-legs due to reflux boiling in the Steam Generator tubes after the loss of natural circulation. The restart of the natural circulation may lead to criticality because of the injection of these low borated slugs towards the core. To evaluate this potential risk, the boron concentration at the core inlet has to be known which makes necessary to estimate the mixing phenomena in the cold leg, in the downcomer and in the lower plenum: CFD calculations are required. First of all the validation of CFX5 CFD code on the relevant phenomena of inherent boron dilution has been established (UPTF TRAM C3 test). Then, an application to the 900 MW French Pressurized Water Reactor series has been performed. (authors)

Science.gov (United States)

There are five tasks within this project on thermally stable coal-based jet fuels. Progress on each of the tasks is described. Task 1, Investigation of the quantitative degradation chemistry of fuels, has 5 subtasks which are described: Literature review on thermal stability of jet fuels; Pyrolytic and catalytic reactions of potential endothermic fuels: cis- and trans-decalin; Use of site specific {sup 13}C-labeling to examine the thermal stressing of 1-phenylhexane: A case study for the determination of reaction kinetics in complex fuel mixtures versus model compound studies; Estimation of critical temperatures of jet fuels; and Surface effects on deposit formation in a flow reactor system. Under Task 2, Investigation of incipient deposition, the subtask reported is Uncertainty analysis on growth and deposition of particles during heating of coal-derived aviation gas turbine fuels; under Task 3, Characterization of solid gums, sediments, and ...

Science.gov (United States)

The largest outbreak of cryptosporidiosis reported in the United Kingdom, involving 575 confirmed cases (of which 474 met an agreed case definition), occurred in the county of Devon during August and September of 1995. The descriptive epidemiology supports the hypothesis that the outbreak was associated with the consumption of cold tap water in the area served by a particular water treatment works. Cryptosporidium oocysts were detected in treated water samples at the time of the outbreak. Although the epidemiological analysis provided strong circumstantial evidence of a waterborne outbreak, the data were not recorded in a manner that made them admissible in criminal proceedings taken by the Drinking Water Inspectorate against the water company involved. The need to carry out an analytical study in conjunction with the identification and characterisation of the ...

International Nuclear Information System (INIS)

The direct injection of steam into a water pool is a method of heat transfer used in many process industries. The amount of research in this area however is limited to the nuclear industry, with applications relating to reactor cooling systems. Electrical resistance tomography (ERT), a low cost, non-invasive and which has high temporal resolution characteristics, can be used as a visualization tool for the resistivity distribution for the steam injection into water pool such as IRWST. In this paper, three dimensional resistivity distribution of the process is obtained through ERT using iterative Gauss-Newton method. Numerical experiments are performed by assuming different resistive objects in the water pool. Numerical results show that ERT is successful in estimating the resistivity distribution for the injection of steam in the water pool

International Nuclear Information System (INIS)

DBEs (Design Basis Event) of KALIMER-600 (Korea Advanced Liquid Metal Reactor) were analyzed in one dimension by KAERI (Korea Atomic Energy Research Institute). KALIMER-600 is the pool type SFR (Sodium cooled Fast Reactor), thereby the sodium of primary system is prohibited movement to out of a reactor vessel. There are many contacting and including compositions in the sodium hot pool, such as IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), Pump, UIS (Upper Internal Structure), and core. Moreover, the complex phenomena are occurred in sodium hot pool during steady and transient states. Therefore, the one dimensional analysis is modified to the multi-dimensional analysis through modification of sodium hot pool from one to three dimensions

Energy Technology Data Exchange (ETDEWEB)

This analysis defines and evaluates the surface water supply system from the existing J-13 well to the North Portal. This system includes the pipe running from J-13 to a proposed Booster Pump Station at the intersection of H Road and the North Portal access road. Contained herein is an analysis of the proposed Booster Pump Station with a brief description of the system that could be installed to the South Portal and the optional shaft. The tanks that supply the water to the North Portal are sized, and the supply system to the North Portal facilities and up to Topopah Spring North Ramp is defined.

International Nuclear Information System (INIS)

Tarapur Atomic Power Station Unit-3 and 4 (TAPS -3 and 4) are the 540 MWe reactors. Unit-4 attained first criticality on 06th March 2005 and operated for about 230 effective full power days (EFPD). Unit-3 attained first criticality on 21st May 2006 and operated for about 20 EFPD. With the reactor operation radiation field increases on the Primary Heat Transport system equipments, Moderator system equipments and auxiliary system equipments due to deposition of fission products and activation products in different reactor systems. These dose rates significantly contributes to the external exposure and stations collective dose during reactor operation, refueling operation and maintenance activities. A study was undertaken at TAPS 3 and 4 to identify the system equipments showing the significant dose rates and identify the radionuclides present in the primary heat transport system, Moderator systems, cover ...

International Nuclear Information System (INIS)

... power reactors pressure tube reactors reactors THERMAL REACTORS.

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 building contained the ...

International Nuclear Information System (INIS)

A design for passive cooling of large containment structures has progressed sufficiently to move forward into the detailed design stage necessary for plant construction. For such application, a safety analysis report has already been submitted to the US Nuclear Regulatory Commission. The design considers an annulus between the inner steel containment vessel and outer, thick-walled concrete shield building with chimney-like natural convection cooling driven only by a density gradient relative to the atmosphere. Air within the annulus is heated as internal containment temperature rises and heat is transferred through the steel containment shell. The resulting air density gradient between the annulus and the environment causes the heated air to rise, producing a natural convection flow through inlets in the shield building, past the steel shell, and out an exit chimney. Several options for enhancing passive heat removal of large containment buildings have been ...

Energy Technology Data Exchange (ETDEWEB)

The WBURN (2-D, 2-group, coarse mesh) code is developed to analyze the equilibrium core characteristics of CANDU-PHWR. The equilibrium characteristics of Wolsung reactor computed by using WBURN are compared with the values given in the Wolsung FSR. The changes of equilibrium core characteristics caused by the variation of design parameters for operating conditions are also investigated. The numerical results indicate that the average discharge irradiation in the Wolsung reactor can be increased up to about 5%.

Energy Technology Data Exchange (ETDEWEB)

Cellulose raw materials costs must be considered in order to obtain a minimized hexose cost. In recognition of this fact, it may be economically advantageous to operate at less than maximum hexose concentration in the reactor and to recycle unreacted cellulose. The objective of this article is to optimize a cellulose-recycle reactor system for producing hexose at minimum cost. A sensitivity analysis of the important variables in the mathematical model of this system is also discussed.

International Nuclear Information System (INIS)

To make a quantitative comparison of risks between the SNR 300 and a modern PWR (Biblis B), the consequences of an accident or the extent of damage of a release of radionuclides to the environment due to an accident are estimated by computer programs for accident consequence models. The accident analysis includes an analysis of events for Bethe-Tait accidents with failure of the outer containment. The FGSB release rates are compared with those of the Society for Reactor Safety (GRS). (HP).

Energy Technology Data Exchange (ETDEWEB)

The Fixed Bed Nuclear Reactor (FBNR) is a small 40 MWe reactor based on the Pressurized Water Reactor (PWR) technology. FBNR is an integrated primary circuit and simple in design. It has the characteristics of being small, modular, proliferation resistant, inherently safe and passively cooled reactor with reduced adverse environmental impact. It utilizes the fuel designed for high temperature reactors operating in a relatively low temperature of PWR environment The 15 mm diameter spherical fuel elements are made of TRISO type microspheres embedded in graphite and cladded by SiC. The coolant flow transfers them from the fuel chamber into the core and become fixed forming a suspended core. Any accident signal will cut off the power to the coolant pump causing a stop in the flow. This results in making the fuel elements fall out of the reactor ...

Energy Technology Data Exchange (ETDEWEB)

The Savannah River Swamp is a 3,020 Ha forested wetland on the floodplain of the Savannah River and is located on the Department of Energy`s Savannah River Site (SRS). Major impacts to the swamp hydrology occurred with the completion of the production reactors and one coal-fired powerhouse at the SRS in the early 1950`s. Water was pumped from the Savannah River, through secondary heat exchangers of the reactors, and discharged into three of the tributary streams that flow into the swamp. This continued from 1954 to 1988 at various levels. The sustained increases in water volume resulted in overflow of the original stream banks and the creation of additional floodplains. Accompanying this was considerable erosion of the original stream corridor and deposition of a deep silt layer on the newly formed delta. Heated water was discharged directly into Pen Branch and ...

Science.gov (United States)

The thermal dissolver, the main reactor of the SRC unit, has suffered a recurring problem. Specifically, it has been observed that whenever the reactor vessel is cooled to below 400/sup 0/F, its bottom head gasket leaks. An analysis of the thermal stress induced in the gasket, owing to transients across the bottom head flange, was sought. The analysis was facilitated by judiciously dividing a symmetric section of the reactor into 79 differential elements. Heat balances have been developed around each element. A numerical technique, the backward finite-difference approach, was employed to obtain the thermal behavior across the bottom head flange as a function of reactor heat-up time. The analysis performed affords an explanation for the failure of the gasket. Based on results of this work, recommendations have been suggested to provide the ...

Energy Technology Data Exchange (ETDEWEB)

This hazard analysis (HA) has been prepared for the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility), in compliance with the requirements of Westinghouse Hanford Company (Westinghouse Hanford) controlled manual WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual, and to the direction of WHC-IP-0690, Safety Analysis and Regulation Desk Instructions, (WHC 1992). An HA identifies potentially hazardous conditions in a facility and the associated potential accident scenarios. Unlike the Facility hazard classification documented in WHC-SD-NR-HC-004, Hazard Classification for 300 Area N Reactor Fuel Fabrication and Storage Facility, (Huang 1993), which is based on unmitigated consequences, credit is taken in an HA for administrative controls or engineered safety features planned or in place. The HA is the foundation for the accident analysis. The ...

Energy Technology Data Exchange (ETDEWEB)

The study required by the West German Ministry of Research and Technology (RS 605) for the Committee on 'Future Nuclear Energy Policy' of the 9th German Parliament is concerned with the following main points: 1) Assessment of technical risks from the social aspect; 2) Discussion of terms and quantification of risks; 3) 'Engineering judgment' and 'questionable' methods in the Fast Breeder analysis of the Society for Reactor Safety (GRS); 4) Assessment criteria of potential damage.

Energy Technology Data Exchange (ETDEWEB)

The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel in 2009. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of control rod ejection transient analysis for these mixed cores

Energy Technology Data Exchange (ETDEWEB)

Condensing two-phase channel flow occurs in many industrial applications, including heating and refrigeration systems. It can also occur in certain nuclear reactor accidents. For example, during a small-break loss-of-coolant accident in a pressurized water reactor, following the partial depletion of the primary coolant, condensation of steam on the primary side of the steam generator tubes can provide a heat sink for disposal of the decay heat generated in the reactor core. Condensing two-phase flow can also play an important role in the operation of the passive emergency cooling system in the advanced simplified boiling water reactor. Here, steady-state condensation in the presence of a noncondensable in a concurrent two-phase channel flow is analyzed using a two-fluid model. The effect of noncondensables on the combined heat transfer at the liquid-gas mixture ...

Energy Technology Data Exchange (ETDEWEB)

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for ...

International Nuclear Information System (INIS)

Understanding of the thermal hydraulic phenomena is now the key issue in solving the severe accident problems of light water reactors. The Atomic Energy Society of Japan has organized a special committee on the evaluation of the thermal hydraulic phenomena in severe accident. The committee has continued the investigation of present status of thermal hydraulics in severe accident. Industries have completed the detailed implementation of the accident management measures, and industries have established also a self-regulatory document mainly on phase II accident management for the containment design of the future reactors. Present paper reviews the current status of evaluation activity referring to severe accident research in Japan. The phenomena included in this paper are (1) molten core behavior in lower plenum of pressure vessel, (2) fuel-coolant interaction, (3) molten core-concrete interaction, (4) direct containment ...