International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The addition of ZnO, depleted in the Zn-64 isotope, to the water of boiling water nuclear reactors lessens the accumulation of Co-60 on the reactor interior surfaces, reduces radioactive wastes and increases the reactor service-life because of the inhibitory action of zinc on inter-granular stress corrosion cracking. To the same effect depleted zinc in the form of acetate dihydrate is used in pressurized water reactors. Gas centrifuge isotope separation method is applied for production of depleted zinc on the industrial scale. More than 20 years of depleted zinc application history demonstrates its benefits for reduction of NPP personnel radiation exposure and combating construction materials corrosion.

International Nuclear Information System (INIS)

In a supercritical water-cooled reactor, property of water changes significantly around the critical point. It is expected that irradiation and change of water property will affect the chemistry and material corrosion. Deep understanding of interactions between supercritical water and materials under irradiation is important. However, comprehensive data on radiolysis, kinetics, corrosion and thermodynamics have not been obtained due to the severe experimental condition. To get such data by experiments and computer simulations, a national program funded by Ministry of Education, Culture, Sports, Science and Technology (MEXT) has been started since December 2002. (author)

International Nuclear Information System (INIS)

This paper presents the objective, scope, and status of the U.S. Department of Energy's Materials and Structures Program to develop a data base on mechanical properties of structural materials for out-of-core structures and components for LMFBRs. Information on the development of a reference data base on materials for the reactor system, reactor enclosure system, primary heat transport system, intermediate heat transport system, and steam generator system is included. In addition, the development of the data and analyses to account for the effects of temperature and stress, as well as water/steam, sodium, and radiation environments, is described. Plans for the development of alternative materials for future out-of-core applications are presented. (author).

International Nuclear Information System (INIS)

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

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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)

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 parameter which is ...

International Nuclear Information System (INIS)

Manganese is a common contaminant of mine water and other waste waters. Due to its high solubility over a wide pH range, it is notoriously difficult to remove from contaminated waters. Previous systems that effectively remove Mn from mine waters have involved oxidising the soluble Mn(II) species at an elevated pH using substrates such as limestone and dolomites. However it is currently unclear what effect the substrate type has upon abiotic Mn removal compared to biotic removal by in situ micro-organisms (biofilms). In order to investigate the relationship between substrate type, Mn precipitation and the biofilm community, net-alkaline Mn-contaminated mine water was treated in reactors containing one of the pure materials: dolomite, limestone, magnesite and quartzite. Mine water chemistry and Mn removal rates were ...

International Nuclear Information System (INIS)

Existing applications of Alloy 800 are summarized, with particular reference to its use in various types of reactor. The need for a co-ordinated research and development programme is stressed, and the variables to be explored are outlined. The papers relating to the problem of corrosion and cracking in water and steam are considered. the strength and ductility of Alloy 800 is considered. Finally, sections of the summary deal with the use of Alloy 800 for (a) sodium cooled fast reactor boiler tubes; (b) the high temperature gas cooled reactor; and (c) PWR steam generator tubes. (U.K.).

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Optimal neutron energy for boron neutron capture therapy (BNCT) has been studied. Epithermal neutron is superior to thermal neutrons in treating deep-seated tumors. Design of the epithermal neutron column for BNCT has been performed by using a two-dimensional transport calculation code. Aluminum and heavy water are used as moderation materials. A thermal neutron column is also designed using heavy water as thermalization material. The configuration of the facility for treatment and research of BNCT and also for basic radio-biological studies of neutrons has been presented.

International Nuclear Information System (INIS)

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility ...

International Science & Technology Center (ISTC)

Development of Ceramic Composite Materials and Structural Elements for High-Temperature Nuclear Reactors

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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 ...

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The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and neutron fluxes, making it the most severely damaged torus component. The collector plate is composed of a protection plate, which is directly exposed to the particle flux, and a heat sink which provides support for the protection plate and carries the water coolant. The high-Z refractory metals have been considered for use as the protection plate material, and austenitic stainless steels and copper alloys have been considered as the heat sink material. Tungsten and Type 316 stainless steels have been selected for the protection plate and heat sink, respectively. The protection plate has a sputtering lifetime of 1.75 y at a 50% duty factor, while the heat sink is expected to last the lifetime of the reactor.

Energy Technology Data Exchange (ETDEWEB)

The integrity of the RPV head and reactor internals was assessed by means of fluid-structural analyses using a coupled method to evaluate the water hammer phenomenon arising from high burnup fuel failure under RIA conditions. The fluid viscosity effect on the water column burst as well as the complex three-dimensional flow paths caused by a core shroud and standpipes were considered in this study. The three analysis scenarios were designed to investigate the above mentioned influential factors separately. In the first scenario, a two-dimensional axisymmetric reactor vessel model without any reactor internals was modeled to assess the influence of the fluid dynamics in the NSC RIA regulatory evaluation. This model has an actual RPV geometry and can be simply separated from other influential factors in order to concentrate only on investigation of the fluid viscosity effect. In the ...

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. The dominating ...

International Nuclear Information System (INIS)

GRR-1 is a 5MW open pool type research reactor with MTR-type fuel elements cooled and moderated by light water with beryllium reflectors at the two opposing sides of the core. A graphite thermal neutron column is adjusted to one side of the core. Six radial horizontal beam tubes are available, of which three contain in-pile collimators for neutron scattering instruments. The reactor is currently out of operation for inspection and refurbishment purposes. The core has been dismantled and the fuel elements are stored in the used fuel storage tank. The GRR-1 inspection and refurbishment plan involves inspection and eventually replacement of the reactor's primary cooling circuit. The health physics procedures to be implemented during inspection of the main water outlet are divided in three stages: a) pool dose rate survey from pool top, b) pool drainage by decreasing ...

International Nuclear Information System (INIS)

Decommissioning of radiological and nuclear installations is for this century the new challenge. One of the performance criteria is the reduction of total quantities of radioactive materials (liquid or solid) arising from dismantling and decontamination of radiological and nuclear installations. In this work we present a new application of the water soluble polymers used as: - flocculation agents in treatment and conditioning process within the management of radioactive liquid materials; - strippable coatings on solid materials based on the water soluble polymers. The parameters of water soluble polymers made in our Institute by radiation processing have been analysed, namely the molecular average weight, composition, and efficiency of utilization of these polymeric materials as well as the content of ash, additives, decontamination factor, ...

Science.gov (United States)

Solid solution-strengthened nickel base alloys have been used for nuclear stream generator tubing in pressurized water reactors since the beginnings of commercial nuclear power. The purpose of this paper is to recap and update the authors understanding of the relationship between processing, resulting structure, and properties for Alloy 600 and to discuss the requirements for optimized performance in both primary and secondary environments. Potential replacement materials and their performance will be discussed. Also discussed is the role and importance of system chemistry, bulk and local, and control and its relationship to performance. A discussion of potential mechanisms of environmentally assisted failure is also discussed.

International Nuclear Information System (INIS)

Zirconium-base alloys are used primarily as fuel cladding material and other core structure material in water cooled nuclear power reactors. Main research achievements and problems about corrosion of zirconium alloys are reviewed; the present theories and challenge are summarized. In the 1980s, great progress had been made towards correlating alloy composition, microstructure and irradiation with corrosion resistance. In the 1990s, main researches are focused on exploring actual mechanism of corrosion, optimizing both alloy composition and microstructure in order to minimize the fuel cycle costs through burnup optimization.

KoreaScience (Korea)

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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.

Energy Technology Data Exchange (ETDEWEB)

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

International Science & Technology Center (ISTC)

Behavior of Nuclear Reactor Fuel Assembly Materials during Their Long-Term Dry Storage

International Nuclear Information System (INIS)

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.

International Nuclear Information System (INIS)

SiC has been considered as a primary candidate material for a first wall component in future fusion reactor because it has been claimed that SiC has excellent high-temperature properties, good chemical stability and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, tritium trapping capacity on the surface of SiC was experimentally obtained at the temperature range of 25-800 deg. C in consideration of tritium trapping to the experimental system. The capacity, which was independent of the water vapor pressure in the gas phase and the temperature, was determined as about 10"6 Bq/cm"2. The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface was quantified at the temperature of 25, 500 and 700 deg. C in consideration of the behavior of tritium trapping at change of experimental condition by the numerical curve ...

International Nuclear Information System (INIS)

Full text of publication follows: Slightly over-stoichiometric lithium orthosilicate pebbles have been selected as one optional breeder material for the European Helium Cooled Pebble Bed (HCPB) blanket. This material has been developed in collaboration of Research Center Karlsruhe and the Schott Glass, Mainz. The lithium orthosilicate pebbles are fabricated from lithium hydroxide and silica by a melting and spraying method in a semi-industrial scale facility. Lithium hydroxide was selected as the precursor since enriched lithium hydroxide is commercially available. The lithium orthosilicate pebbles produced by the process contains oxide phases besides orthosilicate, but it was also found that the oxide phases can be decomposed by annealing at high temperatures. The lithium orthosilicate pebbles produced in this way possesses satisfactory pebble characteristics. Therefore, the authors performed out-of-pile annealing tests using the lithium ...

Energy Technology Data Exchange (ETDEWEB)

In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. ...

Energy Technology Data Exchange (ETDEWEB)

Hydrogen injection into the reactor water has been applied to many BWR power stations. Since hydrogen injected accelerates recombination of oxidant generated by water radiolysis, oxidant concentration, such as dissolved oxygen concentration in reactor water can be reduced. As the result of the reduction of oxidant concentration, Electrochemical Corrosion Potential (ECP) at the surface of structural material can be lowered. Lowered ECP moderates Stress Corrosion Cracking (SCC) sensitivity of structural materials, such as stainless steels. As usual, hydrogen injection system begins to work after the plant start-up is finished, when the condition of normal operation is established. Accordingly, Hydrogen Water Chemistry (HWC) does not cover all the period of plant operation. As far as SCC crack growth is considered, loss of ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

The newly developed techniques by the Hot Laboratory (JMTR HL) have provided for us the key information on behavior of specimens due to mechanical / physical / chemical / synergistic effects of radiation, stress and water for fission and fusion reactor environment. These techniques are focused on several topics as follows; (1) miniaturized specimen test for the development of fusion reactor materials, (2) slow strain rate tensile testing (SSRT) and crack propagation measuring tests for the study of Irradiation Assisted Stress Corrosion Cracking (IASCC) of core internals of LWR, (3) handling technique on specimens including tritium for the research and development of tritium breeders and neutron multiplier as fusion blanket materials, (4) joining method using the Tungsten Inert Gas (TIG) welding technique for re-assembling of capsule and re-fabrication of specimen and (5) ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

Experience at SRS, ORNL, BNL, and Georgia Institute of Technology involving irradiated aluminum clad fuel and target elements, as well as studies of non-irradiated aluminum indicate that some types of aluminum assemblies can be kept in a continually well-deionized water atmosphere for up to 25 years without problems. SRS experience ranges from 2.75 years for the L-1.1 charge kept in deionized D[sub 2]O[sup 1] to greater than 10 years for assemblies stored in the Receiving Basin for Off-site Fuel (RBOF)[sup 2]. Experience at Georgia Institute of Technology reactor in Atlanta yielded the longest value of 25 years without problems. The common denominators in all of the reports is that the water is continually deionized to approximately 2 M[Omega] (2 [times] 10[sup 6]ohms) resistivity and the containers for the water are stainless steel or other non-porous material. This resistivity ...

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)

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)

(1977). United States Parvez, A. Becker, M. Boguslawski, D. Harris,

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 ...

Energy Technology Data Exchange (ETDEWEB)

During 17 years since the start of operation of the first commercial LWR in Japan, many LWRs have experienced various corrosion damages, but the causes of them were clarified, and the counter-measures were executed effectively in actual plants, as the results, the cause of corrosion damage decreased remarkably, and now, the high rate of operation has become to be maintained. In this paper, the major cases of corrosion damage experienced in LWRs in Japan and foreign countries, the causes of them and the countermeasures, the problems of hereafter and so on are described. The corrosion damage of metallic materials in the environment of LWRs occurs in the parts in contact with high temperature, high pressure water and steam, such as stainless steel piping in the primary cooling system of BWRs, and nickel alloy heating tubes of steam generators, carbon steel feed water piping and zirconium alloy fuel cladding tubes in PWRs. ...

International Nuclear Information System (INIS)

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that "2"3"3U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

Briefly described is knowledge of crevice corrosion, corrosion cracking and denting. In evaluating the corrosion resistance of steam generator materials it is necessary to distinguish corrosion problems caused by the primary coolant side and by the secondary circuit side. At present tubes are manufactured of 7 austenitic alloys of a different chemical composition, and available information shows that views on their corrosion resistance differ. Greatest attention has been devoted to corrosion cracking in the presence of NaOH. Findings related to I-600, I-800, I-690 and AISI 316 are given. Corrodibility by sulfur-containing products is now being studied, namely the intercrystalline corrosion cracking caused by the presence of H/sub 2/S/sub 4/O/sub 6/. Knowledge gained in this respect is summed up.

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)

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

International Nuclear Information System (INIS)

The TVR heavy water research reactor was deployed at Moscow Institute of Theoretical and Experimental Physics. In 1990, the final batch of the spent nuclear fuel from this reactor was shipped to Production Association (PA) 'Mayak' for reprocessing. The SNF removal was a stage of the reactor decommissioning activities. The designs of the TVR reactor and its fuel elements are similar to the RA reactor designs. Two ways of the RA reactor SNF transportation to PA 'Mayak' have been considered: in aluminum barrels and in additional canisters using respectively TUK-32 and TUK-19 shipping casks. The practical experience and the equipment used to prepare for the TVR reactor SNF removal can be helpful to the RA reactor personnel in finding the best way to perform these engineering operations. (author)

International Nuclear Information System (INIS)

The application of chemical cleaning for dissolving and removing scale and sludge is being planned in the Japanese pressurized water reactor (PWR) plant in order to maintain high heat transfer performance and to prevent steam generator (SG) tube degradation. In this paper, the effectiveness of the Electric Power Research Institute (EPRI) and German Kraftwerk Union (KWU) processes on the integrity of structural materials other than SG tubes and the comprehensive applicability of chemical cleaning are discussed. The integrity of structural materials such as carbon steel, low-alloy steel and stainless steel was maintained after the EPRI and KWU processes. KWU chemical cleaning tailored for crevice cleaning has been studied to improve its cleaning effectiveness in crevices and to control the corrosion depth of structural materials less than the criterion for corrosion depth. (author)

Science.gov (United States)

average bulk water temper ature rise, OF bulk water temperature at elevation z, OF bulk water temperature in channels 0 and 1, O F film temperature, OF ...

J-STORE (Japan)

Full Text Available

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.

Energy Technology Data Exchange (ETDEWEB)

While tritium exposure to the site-workers in Wolsung NPP is upto about 40 % of the total personnel exposure, Korea Institute of Nuclear Safety has asked tritium removal facility, as one of the requirements for post reactor construction, after operation of four CANDU reactors in Wolsung site. For the purpose of essential removal of tritium from the heavy water system of the heavy water reactors, an experiment of Ar-N{sub 2} cryogenic distillation tower was carried out as a preliminary study for development of liquid-phase catalytic exchange - cryogenic hydrogen distillation process. The steady-state reached after 50 minutes under 90 K in the Ar-N{sub 2} distillation column (inner diameter 20 mm, height 500 mm) packed with Dixon ring ({phi} 3 mm x H 3 mm), and the ratios of Ar-concentration at the top and at the bottom measured by gas chromatography within {+-}1 % relative error was ...

Energy Technology Data Exchange (ETDEWEB)

A method for determining the reactivity of highly subcritical systems of fissile material, using neutron-noise power spectral densities in conjunction with a /sup 252/Cf source, had previousy been tested in two fast reactor critical assemblies (a mockup of the Fast Flux Test Facility reactor and unreflected enriched uranium metal assemblies) and one thermal reactor (a light-water moderated and reflected lattice of Oak Ridge Research Reactor fuel elements). The last-mentioned test demonstrated the effectiveness of the method in water-moderated systems and thereby prompted the present study of its application to facilities for fuel preparation, reprocessing, and storage. To investigate the applicability of this method to facilities for fuel preparation, reprocessing, and storage, limited experiments were performed with a uranyl fluoride ...

Energy Technology Data Exchange (ETDEWEB)

This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial ...

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to date are described for Approved Testing Material (ATM)-103, which is spent fuel from Assembly D101 of pressurized-water reactor Calvert Cliffs, No. 1. This report is one in a series being written by the Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US nuclear waste repository program. ATM-103 consists of 176 full-length irradiated fuel rods with rod-average burnups of about 2600 GJ/kgM (30 MWd/kgM) and less than 1% fission gas release. Characterization data include 1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; 2) isotopic gamma scans; 3) fission gas analyses; 4) ceramography of the fuel and metallography of the cladding; 5) special fuels studies involving ...

Energy Technology Data Exchange (ETDEWEB)

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ...

Energy Technology Data Exchange (ETDEWEB)

Slow, unexpected displacements have been experienced in most pressurized water reactor (PWR) surge lines. Sometimes, these displacement lead to gap closure at the pipe whip restraints. These movements occur because of thermal stratification. This movement has the potential to increase stresses to valves, which may exceed the material yield stress. To understand this phenomenon, Framatome, Commissariat a l'Energie Atomique, and Electricite de France have undertaken large programs for the study of (1) thermal-hydraulic tests with a half-scale Plexiglas surge line, (2) thermal-hydraulic computations of permanent states and transients with a two-dimensional model, and (3) mechanical analysis of displacements and computation of fatigue damage due to stratification. This paper deals with the last subject. Avoiding stratification in piping by process modifications is difficult because of the high flow rate needed. ...

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)

This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capabi...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

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 of uncertainty ...

Energy Technology Data Exchange (ETDEWEB)

Many mines and processing facilities are in remote, difficult to access areas, or are in areas where access is limited or restricted for national reasons. In a joint project with Canada Centre for Remote Sensing, the Canadian Safeguards Support Program is investigating utility of multi- and hyperspectral remote sensing for remotely inspecting such site. With a view to eventually using the hyperspectral satellite data now coming available, airborne data over uranium and copper mine tailings and uranium processing facilities in northern Canada were acquired. The objective of this work is to demonstrate that multispectral and hyperspectral data can provide complementary and supplementary information to high-resolution panchromatic imagery for the following safeguards applications: Inspection aids (up-to-date maps of remote locations), change detection, evaluation of member state declarations, monitoring of reactor, mining and processing facility operations, and ...

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)

The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test ...

Energy Technology Data Exchange (ETDEWEB)

This project of law concerns an additional protocol to the agreement of warranties signed on September 22, 1998 between France, the European atomic energy community and the IAEA. This agreement concerns the declaration of all information relative to the R and D activities linked with the fuel cycle and involving the cooperation with a foreign country non endowed with nuclear weapons. These information include the trade and processing of nuclear and non-nuclear materials and equipments devoted to nuclear reactors (pressure vessels, fuel loading/unloading systems, control rods, force and zirconium tubes, primary coolant pumps, deuterium and heavy water, nuclear-grade graphite), to fuel reprocessing plants, to isotope separation plants (gaseous diffusion, laser enrichment, plasma separation, electromagnetic enrichment), to heavy water and deuterium production plants, and to uranium conversion plants. ...

Energy Technology Data Exchange (ETDEWEB)

The /sup 252/Cf-source-driven neutron noise analysis method has been tested in a wide variety of experiments that have indicated the broad range of applicability of the method. The neutron multiplication factor k/sub eff/ has been satisfactorily detemined for a variety of materials including uranium metal, light water reactor fuel pins, fissile solutions, fuel plates in water, and interacting cylinders. For a uranyl nitrate solution tank which is typical of a fuel processing or reprocessing plant, the k/sub eff/ values were satisfactorily determined for values between 0.92 and 0.5 using a simple point kinetics interpretation of the experimental data. The short measurement times, in several cases as low as 1 min, have shown that the development of this method can lead to a practical subcriticality monitor for many in-plant applications. The further development of the method will require experiments ...

International Nuclear Information System (INIS)

Tarapur Atomic Power Station Unit-4 is first 540 MWe pressurized heavy water reactor in India. It achieved criticality on 06th March 2005 and then operated at full power i.e 500 MWe. Radiation workers during the normal operation and reactor shutdown are exposed to radiation field. The control of dose rates and the collective dose of the radiation workers is most important for the best performance of the reactor. Experience gained during the operation of the 220 MWe reactors has shown that the Moderator system, primary heat transport system, annulus gas system and moderator cover gas system are the main systems contributing to the dose rate and collective dose. In order to identify the radio nuclides contributing to the radiation field, study was undertaken at TAPS Unit-4. Various samples from the Moderator, primary heat transport system, annulus gas system and moderator cover gas ...

International Nuclear Information System (INIS)

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.

Energy Technology Data Exchange (ETDEWEB)

This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). This project has examined the fundamental chemistry of plutonium that affects the integration of hydrothermal technology into nuclear materials processing operations. Chemical reactions in high temperature water allow new avenues for waste treatment and radionuclide separation.Successful implementation of hydrothermal technology offers the potential to effective treat many types of radioactive waste, reduce the storage hazards and disposal costs, and minimize the generation of secondary waste streams. The focus has been on the chemistry of plutonium(VI) in solution with carbonate since these are expected to be important species in the effluent from hydrothermal oxidation of Pu-containing organic wastes. The authors investigated the structure, solubility, and stability of the key plutonium complexes. ...

International Nuclear Information System (INIS)

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

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 ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

Most plans for the disposition of surplus nuclear materials involve storage in sealed containers where the evolution of gases from reactions of adsorbed water could present both pressure and flammability hazards[l] . Despite efforts such as calcining the material to minimize the water content prior to packaging, both residual moisture and readsorbed water may be present in the final containers . Given the anticipated temperature excursions during transportation and storage, this water may thermally desorb, increasing the pressure, and/or thermally dissociate to produce H2 gas, increasing flammability hazards . In addition, the radiation from the nuclear material may induce radiolysis of the water with the likely products being water vapor, H2, 02 and H2O2. In order to better understand the relative ...

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 ...

Science.gov (United States)

The long operating cycle simplified boiling water reactor is a reactor concept that pursues both safety and the economy by employing a natural circulation reactor core without a refueling, a passive decay heat removal, and an integrated building for the reactor and turbine. Throughout the entire spectrum of the design basis accident, the reactor core is kept covered by the passive emergency core cooling system. The decay heat is removed by the conventional active low-pressure residual heat removal system. As for a postulated severe accident, the suppression pool water floods the lower part of the reactor pressure vessel (RPV) in the case when core damage occurs, and the in-vessel retention that keeps the melt inside the RPV is achieved by supplying the coolant. The containment adopts a parallel-double-steel-plate ...

Energy Technology Data Exchange (ETDEWEB)

This patent describes a nuclear system comprising a containment airspace in which a nuclear reactor pressure vessel is disposed there being a reactor core within the pressure vessel. It comprises a heat exchanger elevated a distance above the pressure vessel; a pool of water surrounding the heat exchanger; means for venting the pool of water to an environment outside the containment; a heat exchanger entry conduit within the containment, the entry conduit having an open lower end communicating with the containment space, and an upper end connected to the heat exchanger, water-containing heated fluid present in the containment airspace incident a pressure vessel loss of coolant event entering and flowing through the entry conduit into the heat exchanger for cooling the fluid to convert water vapor therein to a condensate and separate non-condensable gasses ...

Energy Technology Data Exchange (ETDEWEB)

Alloy 690 and Alloy600 are used as a material for the steam generator tubing in the pressurized water reactor(PWR) of nuclear power plants due to its high corrosion resistance. Although those are a highly corrosion resistance material, their stress corrosion cracking(SCC) have been found on occasion, which are deeply related to a surface oxide film on a base material which have occurred on the primary side as well as the secondary side of a tubing. And The SCC is accelerated in the existing Pb which is the impurity of secondary steam generator components. The Oxide on a steel surface in an aqueous solution above 100 .deg. C is composed of a duplex film structure. The inner layer of the oxide is dense and less porous, which is formed by a growth of the oxide layer on the metal surface. The outer layer of the oxide is less adhesive, which is formed by a dissolution and precipitation ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

The status of neutron activation cross sections for some threshold reactions important for reactor materials dosimetry is reviewed. An attempt is made to understand and explain discrepancies between integral and differential data, using recent available experimental results. The importance of standard and benchmark neutron fields for testing differential data for reactor dosimetry is emphasized and the Interlaboratory Reaction Rate (ILRR) program, as well as a similar program pursued by the IAEA, are briefly described.

Energy Technology Data Exchange (ETDEWEB)

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

Energy Technology Data Exchange (ETDEWEB)

One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and ...

Energy Technology Data Exchange (ETDEWEB)

In a reactor environment, the surface of a limiter or wall is primarily determined by the mechanism of erosion and deposition of surface material. It should be possible to use pellet injection to reduce net erosion to zero everywhere if low-Z materials are used for the surface. Erosion rates can, in general, be minimized by large area limiters and high plasma temperatures, which transmit power to the walls with less sputtering. Under ideal steady state conditions the wall surface is dominated by metallurgical effects in the wall.

International Nuclear Information System (INIS)

In an existent emergency reactor core cooling device, if a ruptures should occure in a pipeline of a gravitational dropping type reactor core cooling system pool (GDCS) due to some or other causes, a portion of GDCS pool water was flown out of the ruptured port and could not be used for reactor core cooling. Then, a difference pressure detector is disposed to a GDCS pipeline at the inlet of a reactor pressure vessel. When it is judged by the detector, that coolants flow to the outside of the injection pipeline, an injection value disposed to the GDCS pipeline is closed by the difference pressure signal. Even if a rupture should occur on the side of the pressure vessel at downstream to the check value of the GDCS pipeline, since backflow is caused at the pressure container inlet of the GDCS pipeline with the rupture port, the rupture is detected by the difference pressure detector to ...

Energy Technology Data Exchange (ETDEWEB)

A summary of the off-gas behavior in the HARVEST pot vitrification process is presented. Experimental runs were carried out on 3 representative wastes (MAGNOX - thermal reactor, metal fuel, THORP - thermal oxide fuel and PFR - fast reactor oxide fuel) using 2 methods of feeding the glass-formers (slurry and crizzle). Materials were carried over from the vitrification vessel into the off-gas system by entrainment supplemented by volatilization. The main volatile elements were Ru, B, Cs. Some volatility was also shown by Na and Li. The overall behavior of the off-gas was consistent with the presence in it of 5 separate aerosols of particulate matter. Sources of entrainment gave rise to 3 aerosols, and a further 2 aerosols were formed as a result of chemical reaction (Ru) and condensation (Cs) proceses involving the volatile species. Entrainment was enhanced when the feed contained free alkali nitrate. The Ru volatility ...

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).

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...

International Nuclear Information System (INIS)

Two sodium cooled reactors are currently being operated in the United States of America for the US Department of Energy. These are Experimental Breeder Reactor 11, EBR-11, and the Fast Flux Test Facility, FFTF. EBR-11 is located near Idaho Falls, Idaho, and the FFTF is near Richland, Washington. These reactors are currently engaged in a wide range of testing including fuels and materials tests, and plant system performance and safety development. The US DOE program also includes designs of a next generation sodium cooled power reactor. The FFTF and EBR-11 communities are providing input to these designs. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA.

International Nuclear Information System (INIS)

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

Science.gov (United States)

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

International Nuclear Information System (INIS)

The neutron data required to completely analyze fission reactors includes many isotopes and covers a broad energy range. In both fast and thermal reactors, the neutron inventory is a fine balance determined by the fission properties of "2"3"5U, "2"3"9Pu and "2"3"8U and by the capture cross sections of "2"3"8U, fuel materials, structural materials and coolant materials. In fast reactors, the spectrum of neutrons ranges from 1 keV to 3 MeV and is influenced by the elastic and inelastic scattering properties of "2"3"8U and the structural and coolant materials. For neutron shielding applications, the important neutron data include the total cross sections of structural and coolant materials in the MeV range. The impact of these basic nuclear data in fission reactor applications is most suitably described ...

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)

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

Energy Technology Data Exchange (ETDEWEB)

Steam generators (SG) are the weak link of nuclear power plants, their service life is shorter than the service life of other NPP components. This paper is dedicated to a statistical analysis of SG damages and failures. Heat exchanging tubes (HET) are the most damaged elements in SG, there are on average 286 plugged or repaired tubes in each operating SG. The usually mechanisms of tube failure are the following: denting, corrosion at tube outside, pitting, fretting, and circular crack propagation. Most of damages are located in the transition zone above a tube plate. This study shows that the factors that are involved in the SG HET fault probability are: - design features of SG and secondary equipment elements (high pressure feed heaters (HPFH), low pressure feed heater (LPFH)), - water chemistry at different points of condensate feed pipe, composition and density of deposits on HET surface, efficiency of mechanical and chemical washing, - the physical and chemical ...

International Nuclear Information System (INIS)

The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report ...

International Nuclear Information System (INIS)

The condition of spent nuclear fuel (SNF) in wet storage at ten Soviet-designed research reactors has been assessed in the light of international experience in order to identify any associated safety issues. These reactors use Al-clad UO2-Al or U-Al alloy dispersion fuels of ?20% enrichment that were fabricated in Russia; the reactors have been in operation since 1955-70. Although originally sent for reprocessing, much of the SNF generated over the last 25-30 years has been stored in fuel storage pools (FSPs) of variable water quality. The external condition of wet-stored SNF assemblies from the reactors surveyed varied from significant failure due to galvanic corrosion that was driven by poor water quality, through gradual pitting caused by slightly impure water, to a stable condition of no observable change in the oxidized Al alloy surface ...

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 material temperatures at ...

International Nuclear Information System (INIS)

Current trends in nuclear power generation (and particularly in pressurized water reactors) are toward plant life extension and extended fuel burnup. A higher heat generation rate can induce local boiling regimes at the fuel rod surface in the hottest channels of the core, which can strongly modify the chemical environment of the cladding and influence the oxidation rate of zirconium alloys. Tests performed in out-of-pile loops under severe chemical and thermal-hydraulic conditions (nucleate boiling, higher lithium contents compared to PWRs) reveal two important phenomena: an increase of the oxidation rate of Zircaloy-4 cladding materials in 'high' lithiated environments; an enrichment of the chemical additives in the primary water (boron, lithium) at the surface of the cladding under nucleate boiling conditions. The latter phenomenon, also called 'hideout effect', is mainly controlled by some thermal ...

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.

Science.gov (United States)

In this paper, a novel type of a corona discharge plasma reactor was designed, which consists of needle-plate-combined electrodes, in which a series of needle electrodes are placed in a glass container filled with flue gas, and a plate electrode is immersed in the water. Based on this model, the removal of NO and SO2 was tested experimentally. In addition, the effect of streamer polarity on the reduction of SO2 and NO was investigated in detail. The experimental results show that the corona wind formed between the high-voltage needle electrode and the water by corona discharge enhances the cleaning efficiency of the flue gas because of the presence of water, and the cleaning efficiency will increase with the increase of applied dc voltage within a definite range. The removal efficiency of SO2 up to 98%, and about 85% of NOx removal under suitable conditions is obtained in our experiments.

International Nuclear Information System (INIS)

... 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.

International Nuclear Information System (INIS)

The 105-KE and 105-KW Fuel Storage Basins were constructed more than 35 years ago as repositories for irradiated fuel from the K East and K West Reactors. Currently, the basins contain irradiated fuel from the N Reactor. To continue to use the basins as desired, seismic adequacy in accordance with current US Department of Energy facility requirements must be demonstrated. The 105-KE and 105-KW Basins are reinforced concrete, belowground reservoirs with a 16-ft water depth. The entire water retention boundary, which currently includes a portion of the adjacent reactor buildings, must be qualified for the Hanford Site design basis earthquake. The reactor building interface joints are sealed against leakage with rubber water stops. Demonstration of the seismic adequacy of these interface joints was initially identified as a key issue in the ...

Science.gov (United States)

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

Science.gov (United States)

Record production of heavy water was achieved: the plants at Glace Bay and Port Hawkesbury, Nova Scotia, produced a total of 560 megagrams. A shipment of 500 Mg was delivered on time to the Wolsung CANDU reactor in Korea. Energy conservation and waste hea...

International Nuclear Information System (INIS)

TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U"2"3"5 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different ...

International Nuclear Information System (INIS)

Considerations for longevity of nuclear facilities and ease of decommissioning are of great importance for future nuclear power plants. To this end, a concept of an optimal structural concept for nuclear reactor buildings has been studied: the main feature of this concept is to utilize large-sized, light weight prefabricated members with ultrahigh strength materials. The following two items have been selected to study the prospective structure: (1) Applicability of ultrahigh strength materials for reinforced concrete shear walls (2) Construction using large sized prefabricated members As the first step (1), material and structural tests using ultrahigh strength materials, and the subsequent analysis of those tests for reinforced concrete shear walls, has been conducted. The positive results of this study show a bright future for the use of ultrahigh strength ...

International Nuclear Information System (INIS)

A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization.

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

Energy Technology Data Exchange (ETDEWEB)

The results of an investigation of the effects of several parameters on the reactivity instability of a Boiling Water Reactor (BWR) calculational model are summarized. Calculations were performed for a typical BWR operated at low flow conditions, where reactivity instabilities are more likely to occur. The parameters investigated include the axial power shape (characterized by two separate parameters), the core pressure, and operating flow. All calculations were performed using the LAPUR code which was developed at the Oak Ridge National Laboratory for the dynamic modeling of large BWR's. 4 refs., 8 figs.

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.).

Energy Technology Data Exchange (ETDEWEB)

Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

International Nuclear Information System (INIS)

... Sergio Department of Materials, University of Oxford, Oxford (United Kingdom)

Science.gov (United States)

... Page 26 Ultimate compressive strength versus fly ash/cement limits 41 ... 222 El Ultimate compressive strength versus water/(cement + fly ash) 250 ...

International Nuclear Information System (INIS)

The methods are described of determining natural radionuclides, viz., thorium, uranium, radium, radon, polonium, actinium, and protactinium. The analysis is also described of ores, rocks, minerals, water, air. (M.D.).

International Nuclear Information System (INIS)

The helium-cooled HTGR (High Temperature Gas-cooled Reactor) will take an important position in the global energy strategy. It is expected to supply not only electricity but also high quality thermal energy for various industries and local utilities without exhausting any green house effect gas or acid rain gas. The key R and D issue of the HTGR is economical competitiveness, particularly against light water reactors. Due to the poor heat transfer of the single phase helium, the HTGR's volumetric power density is restricted to tenth of corresponding PWR's value so that increasing the power density by improving heat transfer is strongly desired. The standstill can be broken through by adopting gas-solid suspension medium. Its heat transfer performance is quite excellent. Its heat capacity can be increased drastically without excessive pressurization. Although the thermal radiation is a dominant heat transfer mode in high ...

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 phenomena, design and ...

International Nuclear Information System (INIS)

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. The defect rate was twice that in 1978 but still lower than the two previous years. Methods being employed to detect defects include increasing use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failures by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. (author).

International Nuclear Information System (INIS)

Heat transfer augmentation based on the process intensification concept in heat exchangers and thermal reactors has received much attention in recent years, mainly due to energy efficiency and environmental considerations. The concept consists of the development of novel apparatuses and techniques that, compared to those commonly used today, are expected to bring dramatic improvements in manufacturing and processing, substantially decreasing equipment size, energy consumption, and ultimately resulting in cheaper, sustainable technologies. The objective of this paper was to investigate the heat transfer characteristics of tubular thermal reactor using static mixing technology. Glycerin and water were used as the test fluids and water was used as the heating source. The results for heat transfer rate were strongly influenced by tube geometry and flow conditions.

International Nuclear Information System (INIS)

Criticality was first achieved with the Fast Flux Test Facility (FFTF) a little more than 10 yr ago on February 9, 1980. Although the FFTF was designed and built primarily for testing fuels, materials, and components for the liquid-metal fast breeder reactor program, it has, over its first 10 yr of operation, provided valuable information in many other areas. This paper provides a summary of the contributions to the physics of liquid-metal reactors (LMRs) obtained from operation of and testing in the FFTF, with emphasis on some of the more significant and interesting accomplishments.

International Nuclear Information System (INIS)

The paper provides a brief description of the fuel characterization for Fast Breeder Test Reactor (FBTR) and Prototype Fast Breeder Reactor (PFBR). The development and characterization of mechanical properties of Alloy D9 clad and wrapper tubes are discussed. The problems associated with fusion welding of Alloy D9 are outlined. Non-destructive characterization of cladding tubes by optimum encircling eddy current probes, on-line and off-line neural network methods is presented. Both the on-line and off-line neural network methods could readily detect and size defects specified by the designers

Science.gov (United States)

Following studies on fiber and sheet texture of hexagonal crystal system in 1988, work has been extended to tube texture. Using the zircaloy-4 fuel cladding of Wolsung-type reactor as specimen, six pole figures for different crystallographic planes were m...

International Nuclear Information System (INIS)

An important application of metal hydrides is as a moderator material in nuclear reactors. The fundamental properties of hydrides are illustrated and an impression given of the current research into hydrogen in transition metals. Phase diagrams, magnetic properties, temperature dependence of the diffusion coefficient, energy level schemes and superconductivity are considered. (C.F.).

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)

MOX fuel has been one of the mature nuclear fuels which can be used in light water nuclear power plant now. The development status in this domain in foreign countries, the major influence of MOX fuel on reactor performance and the countermeasures are introduced in this paper. The application of MOX fuel in China's PWR is discussed in the end. (authors)

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

Previous experiments in fast critical assemblies demonstrated a method of determining reactivity from power spectral density measurements with /sup 252/Cf. This method determines reactivity from properties of the reactor only at the subcritical state of interest, thus it does not require a calibration near delayed criticality. The interpretation of the measured data to obtain reactivity does not require knowledge of the relative or absolute values of the source intensity, knowledge of the detection efficiencies, or knowledge of the detection instrumentation frequency responses. An experiment was performed at the Pool Critical Assembly to evaluate the possibility of /sup 252/Cf source driven neutron noise spectral density measurements in light water moderated reactors. This experiment showed that using commercially available detectors, such measurements can be performed in a reasonable time, that is, the measurement of the ...

International Nuclear Information System (INIS)

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

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 the subchannel code was evaluated for a high temperature gas cooled ...

British Library Electronic Table of Contents (United Kingdom)

A dense Pd-Ag membrane reactor (MR) with 100% hydrogen selectivity packed with either Rh/La2O3 or Rh/La2O3-SiO2 as catalysts was used to carry out the dry reforming of methane. The membrane reactor simulation was performed using a well-known reactor model. For this purpose, we employed the equations derived from complete kinetic studies of the dry reforming of methane reaction in connection with both catalysts. In addition, we developed the kinetic equation for the reverse water gas shift reaction (RWGS). The combination of detailed kinetic studies with the measured permeation flux for the Pd-Ag membrane allowed a complete comparison between experimental and simulated operation variables. The variables studied for both catalysts were methane conversion and hydrogen permeation as a function...

International Nuclear Information System (INIS)

This paper describes two research programs which are currently underway in the author's laboratory to investigate the effect of fluid flow on the degradation of power plant materials in high temperature/high pressure aqueous environments. These programs include the design and operation of a controlled hydrodynamic corrosion testing apparatus that can be used to study the general and localized corrosion characteristics of alloys in simulated nuclear reactor environments, and a study of the effect of flow velocity on the stress corrosion cracking of ASTM A508 C1.2 steel and Type 304SS in simulated BWR heat transport fluids.

International Nuclear Information System (INIS)

The objective of this study is to comprehend the basic structural characteristics of box shaped mixed structures proposed for a future nuclear reactor building structure. Specimens of reinforced concrete precast panel walls of the mixed structures were prepared using ultrahigh strength materials. Two bending shear tests were conducted with a parameter of the quantity of reinforcement bars. The results include: (1) Relationship of shear stress and the angle of the structure, and (2) Failure mode. (author)

International Nuclear Information System (INIS)

Plutonium and minor actinides produced as by-products of the UO_2 nuclear cycle could be considered as waste or energy source depending on the strategy selected in the nuclear energy programme. Considering Pu and Minor Actinides as a source, they can be burned in existing water reactor for diminishing the radiotoxicity of the spent fuel, it is necessary to use 'inactive' materials as matrix like ThO_2. ThO_2 matrix has demonstrated its Pu burning efficiency and higher corrosion resistance than UO_2. Uranium-plutonium mixed oxide (MOX) fuel efficiency is low because the presence of U in MOX results in the creation of some new Pu under irradiation. The dissolution behaviour of irradiated (Th,Pu)O_2 pellets with burn-up of 38.8 MWd/kg Th has been studied in carbonated (20 mM HCO_3"-), deionised and granite ground water solution in a hot cell. The dissolution behaviour of Th, Pu, U and Np was studied in ...

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)

International Nuclear Information System (INIS)

Age is a key parameter when deducing the history of plutonium material, i.e. the plutonium produced in the nuclear reactors. This is of vital importance, when a smuggled plutonium sample has been seized and the origin has to be determined. A methodology is described which allows accurately to determine the age of plutonium material by thermal ionisation mass spectrometry using independent parent/daughter relations. This has been demonstrated for Reference Materials of known ages as well as for real samples. The already established method using gamma spectrometry is compared to this. (orig.)

Energy Technology Data Exchange (ETDEWEB)

The CANDU-type nuclear fuel localization project started in 1981, and mass-production system completed in 1987 through the pilot scale demonstration of fuel manufacturing. Since the completion of the mass-production system, about 24,000 fuel bundles (450 ton-U) had been delivered to Wolsung Nuclear Power Plant by the end of 1992, according to the fuel supply contracts with KEPCO. The superiority of KAERI-made nuclear fuel has been demonstrated by having achieved the highest utilization factor in the world in 1992. In 1993, as contracted, 4,824 fuel bundles well fabricated and delivered to Wolsung Nuclear Power Plant. The process improvement, quality control, safety management, safeguards of nuclear materials and various kinds of audits have also been performed in the course for fuel manufacturing. Especially in 1993, the difficulties of the reduction of participating work-force were overcome by improving the manufacturing techniques, and raising the efficiency of ...

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)

A survey is presented of studies dealing with radionuclide X-ray fluorescence analysis (RXFA) of impurities in the air, water, soil, sediments and biological materials, mainly blood, urine, hair, soft tissues and bones. Tabulated are lists of published works on RXFA of the air, water and biological materials. (ES).

International Nuclear Information System (INIS)

Full text: In heavy-water-moderated power reactors, tritium is primarily produced by neutron capture in deuterium nuclei in the moderator and coolant. For CANDU 6 reactors, the estimated steady-state values are #approx# 3 TBq#centre dot#kg"-"1 D_2O in the moderator and #approx# 74 GBq#centre dot#kg"-"1 D_2O in the coolant. Tritium removal is one option available to reactor operators for use in their heavy water and tritium management strategies. The WTRF is designed to remove tritium from tritiated heavy water in each of the four CANDU units at the Wolsung Site, to immobilize the tritium and to store it on site. The detritiation process is based on three steps: the first one (front-end) involves the transfer of tritium from heavy water to deuterium gas; the second one (enrichment) concentrates the tritium in a cryogenic distillation system ...

International Nuclear Information System (INIS)

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

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 ...

British Library Electronic Table of Contents (United Kingdom)

In this work, long-term operation of a pilot scale mixed anaerobic reactor processing crude glycerol and rapeseed meal is discussed. These materials are generated as by-products of biodiesel production. Mixed reactor was operated under mesophilic conditions for the period of 654 days. Total cumulative production of biogas reached 379 m3 (at atmospheric pressure and ambient temperature). Maximum volumetric loading achieved during the operation was 2.17 kg m?3 d?1 for the crude glycerol dose of 2 L. When dosing crude glycerol as a single substrate, average specific production of biogas of 0.76 m3 per L of the g-phase was achieved. The lack of nutrients in the g-phase had to be compensated by an addition of ammonium nitrogen in the form of urea into the reactor. Long term processing of crude ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Full text: The transmutation of nuclear waste to reduce the burden on a geological repository is a relevant topic within the Program of Nuclear Safety Research of the Research Centre Karlsruhe. Several studies have confirmed that a high efficiency of transmutation of actinides is reached in fast neutron spectrum reactor system. Therefore, an important effort is dedicated to the study of transmutation strategies with different fast reactors and their associated technologies. Moreover, in international contexts as Generation IV International Forum (GIF) and Sustainable Nuclear Energy Technology Platform (SNETP), fast reactors are considered in the frame of sustainable development of nuclear energy and reduction of waste. The systems that are currently under investigation, in the frame of the different fuel cycle scenarios, are liquid metal cooled and gas cooled fast reactors as well as Accelerator Driven ...

International Nuclear Information System (INIS)

By its very nature redox potential measurement is suitable for determining the concentration ratio of a stable redox pair through its interaction with a chemically inert electrode surface but not the absolute concentration of a material. The measured redox potentials agree only rarely with those which are easily calculable theoretically. No individual defined stable redox pair is available in power station water. It is therefore not simply possible to measure definable mixed potentials more precisely. For these reasons redox potential measurement in the power station, as also with other types of water, can no longer be regarded as an indicator, by which it can be established whether oxidizing or reducing materials predominate in the water. (orig.).

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)

The determination of the origin of seized nuclear material is important for authorities in the context of the criminal investigation, in order to return the material to its last legal owner and to help preventing any further diversion of material from this source. Origin determination is based on a complex pattern of parameters obtained through analytical measurements. The information required to determine the origin of nuclear materials may be divided into two categories: endogenous information (e.g. age or mode of production of the material) which is self-explanatory; whereas exogenous information (e.g. dimensions, surface roughness, impurities) requires a database to which the parameters can be compared. The Institute for Transuranium Elements has developed methods to determine characteristic parameters like impurities, surface roughness, or microstructural information. ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

It is needless to say that one of the most important roles of the condensate demineralizer in Japanese boiling water reactors (BWR) is to eliminate such impurities during accidental occurrence of sea water leakage from condensate cooling system. Ion exchange resins packed in condensate demineralizer have also been expected to decrease crud, or corrosion products (CP) in condensate water in order to finally reduce activated corrosion products (ACP) in the reactor coolant loop. It is perceived that crud removal ability of a condensate demineralizer has been improved year by year. And we call this phenomenon as `Aging Effect`. Typical property changes of aged cation exchange resin consisted of an increase of water retention capacity and a change of surface texture. Based on these findings, we formulated a new concept and developed new gel type ion exchange resins ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

Alloy 800 and Alloy 600 are well known for their resistance to corrosion in an aqueous medium at high pressure and temperature, for which they have been widely used for more than 3 decades in different structural components of water refrigerated nuclear reactors, especially as material for the steam generator tubes (SG) in these nuclear plants. The SG tubes in the Atucha I and Embalse Nuclear Plants are made with Alloy 800. The speed of corrosion of these materials in a reactor's refrigerant medium, while very small is perfectly measurable and can be described by parabolic or logarithmic type kinetics. In other words this speed is high in the first states of growth during the formation of a protective oxide film but then drops to almost stationary values. One characteristic of these films is the formation of a double layer (or duplex): i) an internal adhering layer, of approximately ...

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 ...

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)

Wastewater from a food-manufacturing plant with a low concentration of organic matter was treated at 37 centigrade in an anaerobic fluidized-bed reactor or in an upflow anaerobic sludge blanket. As the influent TOC (total organic carbon) concentration decreased, the TOC removal efficiency in these reactors decreased from 85% to 65%. The concentration of suspended solids in the effluent could be reduced to 20 mg/l, which corresponded to 7% of that in the influent. The effluent from both reactors was treated aerobically in a fixed-bed reactor. The TOC concentration and optical density of effluent from the aerobic treatment were reduced to 5 mg/l and 0.005, respectively. When the effluent treated anaerobically or aerobically was passed over an activated carbon column, the effluent TOC concentration was reduced to 2 to 3 mg/l. The conductivity in raw wastewater was remarkably reduced on an ion-exchange ...

International Nuclear Information System (INIS)

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-III experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

The main reasons of tube failures in steam generators (SG) are considered. 1.Stress corrosion craining which has 28% of SG (most of them have stainless steel tubes). 2. Corrosion loss of metal, which accounts for 24% of tubes (phosphate corrosion due to addition of PO into water). 3.Denting-peripheral pressing of tubes in the openings of the foundation plates by the corrosive products, which are formed on internal surface of drillings in the foundation plates made of carbon steel. 4.Separation of a plating layer on tube panels. 5.Dratting-corrosion. 6.Metal fatigue. A series of experiments were conducted to study the influence of material selection on tube reliability (stainless steel 304, inconel-600, mone-400, incalloy-800). The problem of increase of SG elements reliability is a complex one and can be solved by direct selection of material, proper control of water-chemical conditions and other ...

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)

Energy Technology Data Exchange (ETDEWEB)

Shield measurements and four cycles of operating experience have shown the design and construction of radiation shields for the Fast Flux Test Facility (FFTF) reactor and plant to be satisfactory. A number of minor shield deficiencies were found and corrected. Most of these were associated with interfaces between components, each of which was satisfactory by itself. Preliminary evaluation of the shield measurements indicates satisfactory agreement with design calculations. Operator doses to date have been quite small, especially when compared to light water reactor experience.

Energy Technology Data Exchange (ETDEWEB)

Recently the real-time neutron radiography system of the Kyoto University Reactor (KUR) has been developed and practically applied to penetrating the side plates of the MTR type reactor fuels and investigation of moving objects. In this paper an application of the KUR neutron TV system to neutron computed tomography (NCT) is described. By using the NTV system, projection data can be acquired in a single measurement and simultaneously the projection image can be observed on a CRT monitor. The Fourier-convolution technique is used to produce the reconstructed image and its image has a good enough quality for revealing water in a small hole of 1.5 mm in diameter.

Energy Technology Data Exchange (ETDEWEB)

Fundamental phenomena of condensation heat transfer at a steam-water interface have been studied related to the thermo-hydrodynamics of the emergency core cooling system for light water reactors. In this study temperature fluctuations near the interface and in the liquid phase were measured using fine thermocouples for a saturated steam-subcooled water co-current stratified two-phase flow in a nearly horizontal rectangular channel, and heat transfer coefficients were determined experimentally. The values of the condensation heat transfer coefficients in this experiment are from 6 to 40 kW/m{sup 2}K. In the regions of high Reynolds numbers, as the steam Reynolds numbers become larger, the average interfacial heat transfer coefficients tend to increase. The corelations of Nusselt numbers were obtained from the heat transfer data. (author)

International Nuclear Information System (INIS)

Within the framework of the EU power plant conceptual study (PPCS), a modular He-cooled divertor concept with integrated pin array (HEMP) is being developed at the Forschungszentrum Karlsruhe. The design goal is to achieve a high heat flux of at least about 10-15 MW/m"2, which is proposed for a near-term reactor model like DEMO. The development and optimization of the divertor concept require a close link between the main issues: design, analyses, materials and fabrication technology, and experiments with feedbacks between them to be accounted for. Design-specific requirements on materials and fabrication issues will be discussed.

International Nuclear Information System (INIS)

Long-term aging of organic materials in reactor containment buildings has become a major issue within the nuclear community. In this article, the status of radiation-aging qualification test requirements in several countries is reviewed, and problems with the current aging methodologies are described. These problems include dose-rate and synergistic effects and environmental synergisms, which have been found for many different polymeric materials. A number of approaches to improved accelerated-radiation-aging tests for prediction of long-term aging behavior are discussed together with their limitations.

Energy Technology Data Exchange (ETDEWEB)

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor, based on the experienced light water reactor (LWR) technology, aiming at effective utilization of uranium resources, high burn-up and long operation cycle and plutonium multiple recycling. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional LWRs. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPC) in 1998, under technical cooperation with three Japanese reactor vendors. In the core design study of the RMWR, several basic core designs with the high ...

Energy Technology Data Exchange (ETDEWEB)

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