International Nuclear Information System (INIS)

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

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 ...

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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 ...

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The structural integrity of the Fuel Test Loop(FTL) in a Korean experimental reactor is evaluated when the FTL, submerged in a water environment, is subjected to a postulated pipe rupture. The analyses are performed under static and dynamic conditions, imposing the thrust force history at each postulated pipe rupture section. Through analysis the following results are found: 1) A double ended guillotine can not be expected based on the toughness of the material, 2) the structural integrity of the chimney surrounding the FTL would not impede the structural integrity by the pipe whip. All analyses are performed by finite element methods.

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

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)

Energy Technology Data Exchange (ETDEWEB)

The objective of Configuration Risk Management Program(CRMP) is to maintain the safety level by assuring the defense-in-depth of nuclear power plant while the configurations are changed during plant operations, especially for the LP/SD. Such a safety purpose can be achieved by establishing the risk monitoring programs with both quantitative and qualitative features. Generally, the quantitative risk evaluation models, i.e., PRA models are used for the risk evaluation during full power operation, and the qualitative risk evaluation models such as safety function assessment trees are used. Through this study, safety function assessment trees were developed.

International Nuclear Information System (INIS)

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 ...

International Nuclear Information System (INIS)

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

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 ...

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 ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was ...

International Nuclear Information System (INIS)

The K Reactor at the Savannah River Site (SRS) shut-down in spring 1988 for maintenance and safety upgrades. Since that time the receiving stream for thermal effluent, Indian Grave Branch and Pen Branch, have undergone a pattern of post-thermal recovery that is typical of other SRS streams following removal of thermal stress. Divesity of fish and aquatic macroinvertebrate communities has increased and available habitats have been colonized by numerous species of herbaceous and woody plants. K Reactor is scheduled to resume operation in 1991 and operate through 1992 without a cooling tower to cool the discharge. It is likely that the reactor will operate at approximately one-third to one-half of full power (800--1200 MW thermal) during this period and effluent temperatures will be substantially lower than earlier operation at full power. Monthly average discharge temperatures at half-power operation will range from ...

Science.gov (United States)

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

Energy Technology Data Exchange (ETDEWEB)

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) nondestructive evaluation using ...

Science.gov (United States)

The Simplified Boiling Water Reactor (SBWR) designed by General Electric is a natural circulation reactor with enhanced safety features for potential accidents. It has a strong coupling between power and flow in the reactor core, hence the neutronic coupling with thermal-hydraulics is specially important. The potential geysering instability during the early part of a SBWR startup at low flow, low power and low pressure is of particular concern. The RAMONA-4B computer code developed at Brookhaven National Laboratory (BNL) for the SBWR has been used to simulate a SBWR startup transient and evaluate its stability, using a simplified four-channel representation of the reactor core for the thermal-hydraulics. This transient was run for 20,000 sec (5.56 hrs) in order to cover the essential aspect of the SBWR startup. The simulation showed that the SBWR startup was a ...

International Nuclear Information System (INIS)

In Japan, the development of next-generation Light Water Reactor has been launched since April 2008. The development program will be completed in 2015. The purpose of development is to cope with the replacement for existing nuclear power plants after 2030 in Japan and the expanding demand for nuclear power in the world; 'Nuclear Renaissance.' The reactor also aims to be global standard at around 2030. The requirements for global standard and domestic users have been investigated through the feasibility study of past 2 years, 2006-2007, and six innovative features or 'Core-Concepts' were established as follows. A) Reactor core system with uranium enrichment above 5% for significant decrease of spent fuel discharge and prominent higher availability B) Long-life materials and innovative water chemistry technologies for 80 years plant lifetime and significant reduction of occupational ...

Energy Technology Data Exchange (ETDEWEB)

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)

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

International Nuclear Information System (INIS)

(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 ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

The Structural Aging (SAG) Program is carried out by the Oak Ridge National Laboratory (ORNL) under sponsorship of the United States Nuclear Regulatory Commission (USNRC). The Program has evolved from preliminary studies conducted to evaluate the long-term environmental challenges to light-water reactor safety-related concrete civil structures. An important conclusion of these studies was that a damage methodology, which can provide a quantitative measure of a concrete structure's durability with respect to potential future requirements, needs to be developed. Under the SAG Program, this issue is being addressed through: establishment of a structural materials information center, evaluation of structural component assessment and repair technologies, and development of a quantitative methodology for structural aging determinations. Progress to date of each of these activities is presented as well as ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

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 ...

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. The effects of material composition, ...

International Nuclear Information System (INIS)

High temperature thermochemistry of thoria-urania fuel for Advanced Heavy Water Reactor was investigated. Oxygen potential development within the matrix and distribution behaviors of the fission products (fps) in different phases were worked out with the help of thermodynamic and transport properties of the fps as well as fission generated oxygen and the detailed balance of the elements. Some of the necessary data for different properties were generated in this laboratory while others were taken from literatures. Noting the behavior of poor transports of gases and volatile species in the thoria rich fuel (thoria-3 mol% urania), the evaluation shows that the fuel will generally bear higher oxygen potential right from early stage of burnup, and Mo will play vital role to buffer the potential through the formation of its oxygen rich chemical states. The problems related to the poor transport and larger retention of fission ...

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)

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 ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

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)

This paper discusses the remote plugging of leaks in inaccessible pipework, with main reference to small leaks which frequently appear in ancillary cooling water circuits of nuclear reactors. Initially developed to cure problems of the pre-stressed concrete pressure vessels of UK reactors, the ZORIC sealant has been used to repair leaking biological shield pipework on six CANDU reactors. ZORIC is based on a water-soluble epoxy resin and an aqueous suspension of a refined mineral clay. This paper describes the evolution of the sealant, the qualification and testing program, and their application to CANDU reactor systems. 2 refs., 6 figs.

International Nuclear Information System (INIS)

On the basis of environmental considerations, a high temperature gas cooled reactor (HTGR) was proposed as the heat source for the Natuna project for CO_2 conversion. To convert CO_2 to useful products, a large amount of high quality water is required for the chemical processes, boilers and other purposes. One LNG production train (maximum of six trains) would produce 0.4 x 10"9 SCF/d of saleable gas and 1.4 x 10"9 SCF/d of CO_2 (in the case of the Exxon process). This CO_2 gas would then be converted to automobile fuel (methane, methanol), which requires a large amount of water. Natural gas from an off- shore gas field is piped to the petrochemical complex on Natuna Island (about 228 km). Natuna is a small island that, apart from sea water, does not have much available water. The desalination process is considered to be the only solution to the water demand ...

KoreaScience (Korea)

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Energy Technology Data Exchange (ETDEWEB)

JRR-3 is a research reactor of 10 MWt output, which attained the criticality in 1962. All the design, manufacture, installation and others of this reactor were carried out by Japanese technologies, except the fuel and heavy water as the moderator and coolant, therefore it is nicknamed Home-made No.1 Reactor. Recently, due to the change in the state of utilizing research reactors and the rise of quality in the utilization, JRR-3 has become to be unable to meet sufficiently the needs of users. The plan of reconstructing the JRR-3 was considered under such situation, and in order to reuse the reactor building, the reactor proper is removed, and an entirely new, high performance, versatile reactor is to be constructed. In this paper, as to the removal works of the JRR-3 reactor proper, the method of ...

Energy Technology Data Exchange (ETDEWEB)

ORNL has conducted a Nuclear Power Options Viability Study for the Department of Energy. That study is primarily concerned with new technology which could be developed for initial operation in the 2000 to 2010 time frame. Such technology would have to compete not only with coal options but with incrementally improved commercial light-water-reactors. This survey reported here was undertaken to gain an understanding of the nuclear commercial technology likely to be offered in the late 1980s and perhaps beyond. The three US vendors actively marketing NSSSs are each developing a product for the future which they expect to be more reliable, more maintainable, more economical, and safer than the present plants. These are all essentially 3800-MW(t) designs, although all are studying smaller plants. They apparently will be off offered as standard prelicensed designs with much larger scope than earlier NSSS offerings, with the possibility of firm prices. Westinghouse with ...

Energy Technology Data Exchange (ETDEWEB)

Studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington have focused on developing and evaluating the reliability of nondestructive testing (NDT) approaches for coarse-grained stainless steel reactor components. The objective of this work is to provide information to the United States Nuclear Regulatory Commission (NRC) on the utility, effectiveness and limitation of NDT techniques as related to inservice testing of primary system piping components in pressurized water reactors. We examined cast stainless steel pipe specimens containing thermal and mechanical fatigue cracks located close to the weld roots and having inner and outer diameter surface geometrical conditions that simulate several water reactor primary piping configurations. In addition, segments of vintage centrifugally cast piping were examined to characterize ...

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)

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

International Nuclear Information System (INIS)

Agricultural practices frequently cause the development of a soil compacted layer below the surface. These compacted layers restrict the root penetration into deeper layers of soil, in search for water. It is proposed to monitor, using Non Destructive Test, the roots growth due to the planting of standard seeds in different agricultural soils, in function of their compactness and humidity. It will be used the neutrons beams derived from an irradiation channel called J-9 of the Reactor Argonauta (IEN/CNEN), so that the neutron radiographic images of the soil-plant system can be obtained. Each root can be evaluated for its ability to penetrate into compacted soil layers; this fact would mean an optimization of agricultural harvests. (author)

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.

International Nuclear Information System (INIS)

Vertical U-tube steam generators in Pressurized Water Reactors (PWRs) operating an All Volatile Treatment (AVT) secondary chemistry have experienced corrosion problems, particularly denting and sludges. The studies reported evaluate the feasibility of using a low-concentration (0.5 wt%) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The process potentially may be applied at schedule intervals, such as during normal refueling outages, to maintain a steam generator in clean operating condition. This report describes the results of testing to evaluate the effectiveness of several chelant acids for dissolving steam generator sludges and crevice magnetite. Corrosion of carbon steel by the chelant acids and the effects of various inhibitors are evaluated. The ...

Energy Technology Data Exchange (ETDEWEB)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens ...

Energy Technology Data Exchange (ETDEWEB)

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

Science.gov (United States)

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

International Nuclear Information System (INIS)

In the case where uranium recovered by an advanced aqueous reprocessing is utilized in light water reactors (LWRs) with the thermal neutron spectrum, the effects of the decontamination factor (DF) of the reprocessing on core neutronic characteristics were examined. The amounts of transuranium (TRU) elements and fission products (FPs) contained in the recovered uranium depend on the DF of reprocessing, and also "2"3"6U is generated by neutron capture of "2"3"5U during a reactor operation. These all act as poisons in the fuel. Therefore, in this paper, the additional "2"3"5U enrichment necessary to compensate for the produced TRU elements, FPs, and "2"3"6U was evaluated for three representative DF values: 10"2, 10"3, and infinity. The low value of 10"2 corresponds to the advanced aqueous reprocessing investigated here. An APWR core with a discharge burnup of 49 GWd/t when the initial "2"3"5U enrichment is ...

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)

Real-time neutron radiography is being evaluated for studying the dynamic behavior of two-phase flow and for measuring void fraction in vertical and inclined water ducts. This technique provides a unique means of visualizing the behavior of fluid flow inside thick metal enclosures. An air-water flow system was constructed to simulate vapor conditions encountered in a fluid flow duct. Air was injected into the bottom of the duct at flow rates up to 0.47 {ell}/s (1 ft{sup 3}/min). The water flow rate was varied between 0 and 3.78 {ell}/min (0 to 1 gal/min). The experiments were performed at the Pennsylvania State University nuclear reactor facility using a real-time neutron radiography camera. With a thermal neutron flux on the order of 10{sup 6} n/cm{sup 2}{center_dot}s{sup {minus}1} directed through the thin duct dimension, the dynamic behavior of the air bubbles was clearly visible ...

International Nuclear Information System (INIS)

Real-time neutron radiography is being evaluated for studying the dynamic behavior of two-phase flow and for measuring void fraction in vertical and inclined water ducts. This technique provides a unique means of visualizing the behavior of fluid flow inside thick metal enclosures. An air-water flow system was constructed to simulate vapor conditions encountered in a fluid flow duct. Air was injected into the bottom of the duct at flow rates up to 0.47 ell/s (1 ft"3/min). The water flow rate was varied between 0 and 3.78 ell/min (0 to 1 gal/min). The experiments were performed at the Pennsylvania State University nuclear reactor facility using a real-time neutron radiography camera. With a thermal neutron flux on the order of 10"6 n/cm"2#centre dot#s"-"1 directed through the thin duct dimension, the dynamic behavior of the air bubbles was clearly visible through 5-cm (2-in.)-thick ...

International Nuclear Information System (INIS)

Short-lived fission gas release from fuel pellets during irradiation was investigated based on the experimental results of the gas-flow rigs irradiated in the Halden Heavy Water Reactor (HBWR). The release-to-birth (R/B) rates of short-lived fission gas were measured by means of gas-flow measurement during the irradiation experiments. Surface-to-volume (S/V) ratios of fuel pellets and diffusion coefficients of short-lived fission gas release were evaluated from the obtained (R/B) values. The increase of (S/V) ratio agreed well with the point where the fuel temperature exceeded the threshold of 1% fission gas release. This indicates that the interlinkage of fission gas bubbles occurred there. The evaluated diffusion coefficients scattered in the range between 10"-"2"3 and 10"-"1"7 m"2/s, and the effects of fuel type (UO_2 or MOX) were not clearly observed. In addition, it is likely that the restructuring ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

The concept of preliminary transmutation of minor actinides before placement to the long-term storage is considered. The purpose of such preliminary transmutation before ultimate storage is to incinerate a part of actinides and to transform another part into new actinides providing low level of radiotoxicity accumulated in the storage. Modes of transmutation in reactors of PWR, PHWR (CANDU), and Superfenix types are compared. Among power reactors, heavy-water PHWR type reactor is most acceptable for preliminary transmutation. (author)

International Nuclear Information System (INIS)

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

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 directly related to ...

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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 ...

International Nuclear Information System (INIS)

Mitsubishi is now developing a next-generation Pressurized Water Reactor (PWR) which has the innovative feature of hybrid safety systems (optimum combination of passive safety system and active safety system) and passive core cooling by horizontal steam generators during Loss of Coolant Accident (LOCA). In order to confirm the capability of this passive core cooling system during LOCAs, the thermal-hydraulic tests of horizontal steam generator and the integral thermal-hydraulic tests simulating the LOCAs were performed. The thermal-hydraulic tests of horizontal steam generator consist of a single tube test and a multi-tubes test. On the basis of these test results, the heat transfer characteristics of steam-water two-phase flow with noncondensable gas along a long horizontal tube is understood and the heat transfer correlation including the effect of noncondensable gas is presented. The integral thermal-hydraulic tests ...

Energy Technology Data Exchange (ETDEWEB)

The most relevant aspects of research and service experience with environmentally-assisted cracking (EAC) of carbon (C) and low-alloy steels (LAS) in high-temperature (HT) water are reviewed, with special emphasis on the primary pressure boundary components of boiling water reactors (BWRs). The main factors controlling the susceptibility to EAC under light water reactor (LWR) conditions are discussed with respect to crack initiation and crack growth. The adequacy and conservatism of the current BWRVIP-60 stress corrosion cracking (SCC) disposition lines (DLs), ASME III fatigue design curves, and ASME XI reference fatigue crack growth curves, as well as of the GE EAC crack growth model are evaluated in the context of recent research results. The operating experience is summarized and compared to the experimental/mechanistic background knowledge. Finally, open ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The verification test program on zinc injection applicability to Japanese BWRs was started in 1997. Laboratory tests using high temperature water loops under BWR reactor water conditions are in progress. This paper is an interim report on results obtained so far. Co-58 and Zn-65 were simultaneously used in the Co radioactivity buildup test to evaluate zinc injection suppression effects towards cobalt deposition on pre-oxidized stainless steel. The following results were obtained. The Co deposition was suppressed effectively by Zn injection, even when there was a pre-oxide film. For the test piping that had the pre-oxide film formed under the NWC (normal water chemistry) condition, when soaked under the HWC (hydrogen water chemistry) condition a large amount of Co-58 was taken into a small part of the inner layer. The distribution ratio of Co-58 in the inner ...

Energy Technology Data Exchange (ETDEWEB)

Chemical-looping combustion is an indirect combustion technology with inherent separation of the greenhouse gas CO{sub 2}. The feasibility of using NiO as an oxygen carrier during chemical-looping combustion of coal has been investigated experimentally at 800-960{degree}C in the present work. The experiments were carried out in a fluidized bed, where the steam acted as the gasification-fluidization medium. Coal gasification and the reaction of oxygen carrier with the water gas take place simultaneously in the reactor. The oxygen carrier particles exhibit high reactivity above 900{degree}C, and the dry basis concentration of CO{sub 2} in the exit gas of the reactor is nearly 95%. The flue gas composition as a function of the reactor temperature and cyclic reduction number is discussed. At 800-960{degree}C, the dry basis concentration of CO{sub 2} in the flue gas presents a monotonously increasing trend, ...

Energy Technology Data Exchange (ETDEWEB)

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

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 ...

British Library Electronic Table of Contents (United Kingdom)

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25mx0.05m) and 2.59m, respectively, whereas the inclination angle of the riser is 50degree. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels mea...

Energy Technology Data Exchange (ETDEWEB)

The mission of the Heavy Water portion of D Area (or 400 Area) at SRS is to purify the site inventory of heavy water for storage in the Reactor Areas for future DOE missions.

Energy Technology Data Exchange (ETDEWEB)

Since January, 1981, the project of development of nuclear fuel fabrication technology for Wolsung reactor (CANDU type) was undertaken by KAERI(Korea Advanced Energy Research Institute) and successfully fulfilled with loading 24 fuel bundles made by KAERI in Wolsung reactor in September, 1984. On the basis of this accumulated technology and experience, mass production plan to supply all the nuclear fuels for Wolsung reactor is under way. In this presentation, the Korean experience in the development of the nuclear fuel fabrication technology, safety and performance evaluation of KAERI fuel and the results of irradiation of KAERI fuels in Wolsung reactor will be described.

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 ...

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-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)

Energy Technology Data Exchange (ETDEWEB)

The objective of the Hanford Environmental Dose Reconstruction Project (HEDR) is to estimate the potential radiation doses received by people living within the sphere of influence of the Hanford Site. A potential critical pathway for human radiation exposure is through the consumption of waterfowl that frequent onsite waste-water ponds or through eating of fish, shellfish, and waterfowl that reside in/on the Columbia River and its tributaries downstream of the reactors. This document summarizes information on fish, shellfish, and waterfowl radiation contamination for samples collected by Hanford monitoring personnel and offsite agencies for the period 1945 to 1972. Specific information includes the types of organisms sampled, the kinds of tissues and organs analyzed, the sampling locations, and the radionuclides reported. Some tissue concentrations are also included. We anticipate that these yearly summaries will be helpful to individuals and ...

International Nuclear Information System (INIS)

The objective of the Hanford Environmental Dose Reconstruction Project (HEDR) is to estimate the potential radiation doses received by people living within the sphere of influence of the Hanford Site. A potential critical pathway for human radiation exposure is through the consumption of waterfowl that frequent onsite waste-water ponds or through eating of fish, shellfish, and waterfowl that reside in/on the Columbia River and its tributaries downstream of the reactors. This document summarizes information on fish, shellfish, and waterfowl radiation contamination for samples collected by Hanford monitoring personnel and offsite agencies for the period 1945 to 1972. Specific information includes the types of organisms sampled, the kinds of tissues and organs analyzed, the sampling locations, and the radionuclides reported. Some tissue concentrations are also included. We anticipate that these yearly summaries will be helpful to individuals and ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

Tritium, a beta emitting gas at room temperature causes a biological hazard in the locations where it is present beyond acceptable limits. The hazard can be due to inhalation, and absorption by skin. Hence is the necessity of Tritium monitoring instruments/systems for ensuring safety in the PHWRs and the nuclear research plants and laboratories. It is desirable that the instruments address satisfactorily to certain factors like the following: (i) Wide range of Tritium concentrations - 1 to 104 DAC ( Derived Air Concentration) (ii) On-line monitoring features (iii) Small response time (On-spot instantaneous measurements) (iv) Portability (v) Mitigation of memory effects. This paper presents an overview of the Online Tritium in Air Monitoring Systems manufactured by ECIL for Pressurised Heavy Water Reactors at Tarapur, Kaiga, and Rawatbhata. Significant aspects of design, function, testing, limitations of the detectors and electronics and the ...

International Nuclear Information System (INIS)

The TRAC/PFI-MOD2 Version 5.3 code was assessed against six FLECHT-SEASET forced reflood tests (31504, 31203, 31302, 31701, 34209, and 31922) and two cylindrical core test facility (CCTF) tests [C1-19 and C2-6]. The objective of this study was to evaluate the clad thermal response predictive capabilities of the code with the newly added reflood model under large-break loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (PWR). The TRAC model for the FLECHT-SEASET test facility was developed from a RELAP5 model. The test section was modeled using a vessel component with 23 axial levels, 1 radial ring, and 1 azimuthal cell. Test inlet and exit conditions were modeled using fill and break components, respectively. The measured lower and upper plenum test conditions were input to the model. The electrically heated rod was modeled using a rod component with 22 axial mesh points. The axial boundary of each ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The development of crud removal technology has started with the finding of the resin aging effect that an old ion exchange resin, aged by long year of use in the condensate demineralizer, had an enhanced crud removal capability. It was confirmed that some physical properties such as specific surface area and water retention capacity were increased due to degradation caused by long year of contact with active oxygens in the condensate water. So, it was speculated that those degradation in the resin matrix enhanced the adsorption of crud particulate onto the resin surface, hence the crud removal capability. Based on this, crud removal resin with greater surface area was first developed. This resin has shown an excellent crud removal efficiency in an actual power plant, and the crud iron concentration in the condensate effluent was drastically reduced by this application. However, the cross-linkage of the cation resin had to be lowered in a ...

Energy Technology Data Exchange (ETDEWEB)

This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart ...

International Nuclear Information System (INIS)

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized ...

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.

Energy Technology Data Exchange (ETDEWEB)

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

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)

Tricking filters are a very promising alternative for the post treatment of effluents from UASB reactors treating domestic sewage,especially in developing countries. Although a fair amount of information is already available regarding organic mater removal in this combined system, very little is known in relation to nitrogen and surfactant removal in trickling filters post-UASB reactors. Therefore, the purpose of this study was to evaluate and compare the effect evaluate and compare the effect of different application rates and packing media types on trickling filters applied to the post-treatment of effluents from UASB reactors, regarding the removal of ammonia nitrogen and surfactants. (Author)

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)

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.).

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Problem areas in BWRs, PWRs and PHWRs, from the viewpoint of chemistry, and the problem of fission product release in nuclear reactors are discussed. These problem areas are : fuel performance, off-normal water chemistry due to condenser leaks, the transport and deposition of the activated corrosion and fission products, denting in steam generators (in the case of PWRs), ingress of air in the cover gas helium and consequent radiolysis of D_2O in the moderator circuit (in the case of PHWRs). (M.G.B.).

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to ...

UK PubMed Central (United Kingdom)

Data on raw water quality, disinfection treatment practices, and the resulting mutagenic properties of the treated water were compiled from pilot- and full-scale treatment experiments to evaluate that...Full Text Available

International Nuclear Information System (INIS)

1984. p. 890. Germany Picat, P. Thirion, FL Quinault, FM CEA Centre

British Library Electronic Table of Contents (United Kingdom)

In the Mediterranean arc, vines for wine production are mainly grown without the support of irrigation. Under such conditions, site variables affecting the extent and seasonal timing of water deficits are the dominant environmental constraints for grape production. Moreover, water availability and vine water status are the factors most comprehensively determining fruit composition and, thus, wine quality. Therefore, monitoring the extent of water stress in vines might be a valuable tool for the optimisation of grape yield and quality. The objective of this study was to evaluate the feasibility of using the reflectance based Water Index (WI) to estimate vine water status at the leaf and canopy levels. The study was conducted on Vitis vinifera cv. Chardonnay potted plants submitted to contra...

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

The humidification dehumidification desalination process is viewed as a promising technique for small capacity production plants. The process has several attractive features, which include operation at low temperature, ability to utilize sustainable energy sources, i.e. solar and geothermal, and requirements of low technology level. This paper evaluates the characteristics of the humidification dehumidification desalination process as a function of operating conditions. A small capacity experimental system is used to evaluate the process characteristics as a function of the flow rate of the water and air streams, the temperature of the water stream and the temperature of the cooling water stream. The experimental system includes a packed humidification column, a double pipe glass condenser, a constant temperature water circulation tank and a chiller for cooling ...

International Nuclear Information System (INIS)

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

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.

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

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)

The 3001 Storage Canal is located under portions of Buildings 3001 and 3019 at Oak Ridge National Laboratory (ORNL) and has a capacity of approximately 62,000 gallons of water. The term canal has historically been used to identify this structure, however, the canal is an in-ground reinforced concrete structure satisfying the regulatory definition of a tank. From 1943 through 1963, the canal in Building 3001 was designed to be an integral part of the system for handling irradiated fuel from the Oak Ridge Graphite Reactor. Because one of the main initial purposes of the reactor was to produce plutonium for the chemical processing pilot plant in Building 3019, the canal was designed to be the connecting link between the reactor and the pilot plant. During the war years, natural uranium slugs were irradiated in the reactor and then pushed out of the graphite matrix into the system of ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs.

Science.gov (United States)

This report contains the study on the dynamic characteristics of Wolsung fuel rod and on the dynamic balancing of rotating machinery to evaluate the performance of nuclear reactor components. The study on the dynamic characteristics of Wolsung fuel rod wa...

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)

Purpose: To provide a feedwater control device capable of minimizing the adverse response of steam drum level at low power. Consitution: In order to perform feedwater control at low power by the substantial control of three factors, that is, main steam flow rate, feedwater flow rate and steam drum level, the main steam flow rate is determined from the reactor output and feedwater rate is determined from the changes in the feedwater temperature due to the mixing of waters in the reactor clean up system and feedwater. If a difference is resulted between these flow rates, a starting feedwater regulator is controlled instantly to eliminate the difference. The water level in the steam drum is used for amending the difference from the final set value of the drum water level, by which the adverse response of the steam drum level can be minimized. (Seki, T.).

Energy Technology Data Exchange (ETDEWEB)

Progress reports are presented for the following two areas: catalytic cracking studies with water-wet silica-alumina catalysts; and Fischer-Tropsch reactor studies where similarities and differences between fixed bed and slurry type reactors are investigated and further experiments conducted to measure mass transfer coefficients and reaction kinetics which are to be used in a model slurry reactor. The following are some of the conclusions. (1) The premise that the presence of liquid water might increase catalytic cracking activity was found to be invalid. It was demonstrated that cracking can occur at previously unobserved low temperatures (though at low conversions) and that an anomaly exists in that one of the catalysts tested shows an entirely different cracking behavior and probably follows a different cracking mechanism. (2) the diameter of a fixed-bed Fischer-Tropsch ...

International Nuclear Information System (INIS)

Results of calculational investigations into circulation of water and steam-and-water mixture in the PGV-1000 steam generator heat exchanger bundle used at NPPs with the WWER-1000 reactors, are considered. Model of water circulation in horizontal steam generator with submerged heating surface under conditions of steam generation irregularity along the heat transfer tubes is made. On the basis of the obtained data the assumption is made about water essential overflows from the hot collector zone into the cold one. Overflow rate over the upper line of the heat transfer tubes may constitute 0.7 m/s. The conclusion is made about the necessity to set up the vertical barrier which divides hot and cold sections of heat transfer tubes and helps to avoid water transverse overflows.

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)

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 analysis have shown ...

International Nuclear Information System (INIS)

In an attempt to obtain basic data for evaluation of exposure doses in Niigata Prefecture, the concentrations of tritium in atmospheric water, precipitation, river water, and tap water were measured. Samples of atmospheric water, river water, and tap water were collected once for 2 weeks; and precipitation collected for one week was used as sample. The concentration of atmospheric tritium depended on the concentration of tritium in moisture and the content of water in atmosphere. Tritium levels were high in May, November and March, and low in September. Regarding tritium concentrations, there was a good correlation between atmospheric water and precipitation. Tritium concentrations in both of them varied from sample to sample. The concentrations of tritium in river and tap water ...

International Nuclear Information System (INIS)

The CO laser is superior in the absorption characteristic to materials to the CO2 laser due to its shorter wavelength. In consideration of this characteristic Nuclear Power Engineering Corporation is studying this applicability sponsored by the Ministry of International Trade Industry of Japan to cutting of reactor core internals of commercial nuclear power plant. In decommissioning of reactor core internals it is necessary to cut stainless steel plates of 305 mm thick. The authors cut stainless steel plates of up to 310mm thick in air and those of up to 150 mm thick underwater with a 20kW class laser. Further, models simulating key structural elements of PWR core internals were cut and secondary products to clarify the applicability of the CO laser cutting to reactor core internals were evaluated. (author)

Energy Technology Data Exchange (ETDEWEB)

An estimate of the tritium dose to the public in the vicinity of the heavy water research reactor facility at AECL-Chalk River Laboratories, Ontario, Canada, has largely been accomplished from analyses on regularly-collected samples of air, precipitation, drinking water and foodstuffs (pasture, fruit, vegetables and milk) and environmental dose models. To increase the confidence with which public doses are calculated, tritium doses were estimated directly from the ratio of tritiated species in urine samples from members of the general public. Single cumulative 24 h urine samples from a few adults living in the vicinity of the heavy-water research reactor facility at Chalk River Laboratories, Canada were collected and analysed for tritiated water and organically bound tritium. The participants were from Ottawa (200 km east), Deep River (10 km west) and Chalk ...

International Nuclear Information System (INIS)

Decision-making is strongly based on modelling in two of our water supply projects concerning artificial ground water. The model used is a 2-dimensional flow and transport model using the element method. Due to sharp variations in Finnish soils, the modelling process is based on careful and detailed evaluation of basic hydrogeological data before and during model calibration.

Energy Technology Data Exchange (ETDEWEB)

The Oklo natural fossil fission reactors in Gabon, Equatorial Africa, have been studied as a natural analogue for spent nuclear fuel in a geological environment. For these studies, it is important to know what has happened to these reactors since they formed. This review is focussed on existing geological and geochronological information concerning the Oklo reactors and the surrounding ore. A sequence of geological and geochemical events in the Oklo area, as described in the literature, is given. The data and the studies behind this established geochronology are discussed and evaluated. Of the regional geology, special attention is given to the dating of the Francevillian sediments, and the intrusion of a dolerite dyke swarm. The processes that led to the mineralisation at Oklo, the subsequent formation of the nuclear reactors and later migration of fission products are described. ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

The dose from heavy ion beams has been calculated using a one-dimensional transport theory and evaluated for 670 MeV/amu /sup 20/Ne beams in water. The result is presented so as to be applicable to arbitrary ions for which the necessary interaction data are known. The present evaluation is based on the Silberberg-Tsao fragmentation parameters augmented with light fragment production from intranuclear cascades, recently calculated nuclear absorption cross sections, and evaluated stopping power data. Comparison with recent experimental data obtained at the Lawrence Berkeley Laboratory reveals the need for more accurate fragmentation data.

Energy Technology Data Exchange (ETDEWEB)

Public Law 95-209 includes a requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on the safety research program of the Nuclear Regulatory Commission. This report presents the results of the ACRS review and evaluation of the NRC safety research program for Fiscal Year 1983. The report contains a number of comments and recommendations.

Science.gov (United States)

The results of research on the evaluation of the half-bead weld repair method for use on nuclear reactor components are reviewed from data obtained on thick-section test pieces and intermediate-size pressure vessels. Material properties, the magnitude of residual stresses and the structural behavior of flawed pressure vessels are being obtained to determine the adequacy of the weld repair method for application in thick-section components.

International Nuclear Information System (INIS)

In General Electric's design of the Simplified Boiling Water Reactor (SBWR), The depressurization valves (DPV) are installed in the reactor pressure boundary: four are connected to the reactor vessel by means of nozzles, and two more are located on the main steam pipes (one DPV for each line), which act during particular transients and/or loss of coolant accidents (LOCA), consequently providing the reactor vessel with a safe quick depressurization system. Once the vessel is de pressurised, the passive gravity-driven cooling system (GDCS) starts to operate, permitting the injection of water required for continuous core cooling. DPVs are leak tight, with welded flaps, actuated by a [striker[hammer***] which is activated by an explosive mixture. The dynamic loads that open these valves include, in addition to those produced by steam (typical in any thermodynamic ...

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)

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 ...

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 ...

Science.gov (United States)

The effects of variable hardness, pH, alkalinity, humics, and suspended clay on the chemical speciation of copper and its toxicity to fathead minnow larvae in Lake Superior water were investigated. Two proposed methods (toxicity factors and chemical speciation) for predicting LC50 values in specific natural waters from laboratory toxicity data and the average site specific values of general water quality parameters were evaluated. The accuracy of the cupric ion-selective electrode in determining CU/sup +2/ activities in ambient and chemically altered Lake Superior water was also determined.

Energy Technology Data Exchange (ETDEWEB)

In order to realize improve of reliability and economy by duplicate production, rapid supply of repair parts from standardized storage, such were expected as to have continuous order of standardized plant, to ignore site condition, to avoid expansion of regulatory requirement. Standardization program was planned to limitedly promote standardization of safety-related design concept, major specification and basic system composition of reactor and primary systems. The area of standardization had been tried to expand to BOP such as general arrangement and rad-waste system.

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).

Energy Technology Data Exchange (ETDEWEB)

This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

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.

British Library Electronic Table of Contents (United Kingdom)

In this paper we provide a geochemical investigation on 34 groundwater samples in the Mt. Vulture volcanic aquifer representing one of the most important groundwater resources of the southern Italy pumped for drinking and irrigation supply. The present study includes the first data on the abundance and mobility of minor and trace elements and the thermodynamic considerations on water-rock interaction processes in order to evaluate the conditions of alkali basalt weathering by waters enriched in magma-derived CO2. The results highlight the occurrence of two hydrofacies: bicarbonate alkaline-earth and alkaline waters deriving from low-temperature leaching of volcanic rocks of Mt. Vulture, and bicarbonate-sulfate-alkaline waters (high-salinity waters) related to prolonged water circulation in...

International Nuclear Information System (INIS)

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"2K. 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)

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 ...

International Nuclear Information System (INIS)

The CATHENA (formerly ATHENA) has been used to simulate the thermalhydraulic behaviour of the WOLSUNG-1 CANDU-600 reactor during the D_20 spill incident of 1984 November 25. A 4-inch (nominal) Liquid Relief Valve inadvertently opened in the reactor auxiliary system during normal reactor operation, resulting in a discharge of heavy water from the primary heat transport system. The valve remained open for approximately 29 minutes. CATHENA is an advanced thermalhydraulic computer code for analysis of postulated loss-of-coolant accidents (LOCA) and transient faults in CANDU nuclear reactors. A full two-fluid (six-equation) representation of the two-phase flow is used. Component models are used to represent pumps, valves, critical discharge, etc., which are necessary to describe the behaviour of the CANDU system under upset conditions. Heat transfer between the fluid and piping walls (or ...

Energy Technology Data Exchange (ETDEWEB)

To get a considerable reduction of the radiotoxicity due to americium, a thermal neutron fluence of 2.10{sup 22} cm{sup -2} or a fast neutron fluence of 2.10{sup 24} cm{sup -2} is required. Irradiation in a thermal neutron flux leads to lower masses of {sup 234}U and precursors and of {sup 237}Np and precursors, but to higher curium masses and much higher neutron emission rates than irradiation in a fast neutron flux. Therefore, irradiation in a fast neutron flux has preference when multiple recycling is adopted. When once-through burning is applied, irradiation in a thermal neutron flux can be applied. Then irradiation in a heavy water reactor (HWR) has preference above irradiation in a PWR or in a high temperature gas-cooled reactor (HTGR). (authors) 4 refs.

International Nuclear Information System (INIS)

The PANDA test facility at the Paul Scherrer Institute is used to study the long-term performance of the Simplified Boiling Water Reactor's passive containment cooling system. The PANDA tests demonstrate performance on a larger scale than previous tests and examine the effects of any non-uniform spatial distributions of steam and non-condensable gases in the system. The facility is in 1:1 vertical scale and 1:25 scale for volume, power etc. Extensive facility characterization tests and steady-state passive containment condenser performance tests are presented. The results of the base case test of a series of transient system behaviour tests are reviewed. The first PANDA tests exhibited reproducibility, and indicated that the Simplified Boiling Water Reactor's containment is likely to be favorably responsive and highly robust to changes in the thermal transport patterns. (orig.).

Energy Technology Data Exchange (ETDEWEB)

A fusion reactor (such as NET/ITER) which breeds its own tritium fuel requires tritium recovery, purification and separation from the other isotopes. Cyclic adsorption processes are strong candidates for several of the processes involved: amongst other advantages, they promise a low tritium inventory. A good adsorbent for such processes must have high adsorption capacity, high selectivity and very low tritium retention after each cycle. Pure zeolite powder is shown to have an excellent combination of these three properties. However, in practice problems can arise from tritium which is not removed by reactivation. In this paper we show that tritium retention in zeolites can be caused either by water retained in the zeolite structure, which can be removed by ore rigorous activation, or by water tapped on binders in commercial pellets. (orig.).

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

A three-dimensional, continuous-energy, coupled neutron-gamma Monte Carlo model of the advanced neutron source (ANS) final preconceptual reference core design has been developed using MCNP Version 3b. This model contains the reactor core with control rods, the heavy-water reflector tank with shutdown rods and representative beam tubes, and the outer light-water poor. Eighty homogenized fuel zones per fuel element are used to represent the radical and axial {sup 235}U fuel grading. This model is the most sophisticated, physically accurate reactor physics model of the ANS currently available. The purpose of this summary is to demonstrate the MCNP methods and applications for the ANS.

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

AREVA is the world leader in the design and construction of nuclear power plants, the manufacture of heavy components, and the supply of nuclear fuel and nuclear services such as maintenance and inspection. The Equipment Division provides the widest range of nuclear components and equipment, manufactured at its two facilities in Jeumont, northern France, and St. Marcel, in Burgundy. The St. Marcel plant, set on 35 ha (87.5 acres) near Chalon-sur-Saone, was established in 1973 in a region with a long history of specialized metalworking and mechanical activities to meet the demand for non-military nuclear requirements in France. The site offers two advantages: - excellent facilities for loading and transporting heavy components on the Saone river, - it's proximity to other group sites. Since its completion in 1975, the Chalon/St. Marcel facility has manufactured all the heavy components for French pressurized water reactors (PWRs) ...

International Nuclear Information System (INIS)

During 1993 and 1994 a working group of biologists operating in Region Lombardia has carried out a study to evaluate the reliability of artificial substrates in the assessment of water quality by the Extended Biotic Index. Macroinvertebrate samples were collected by means of hand net and artificial substrates (up to 3 replicates) in 22 sampling sites of 15 watercourses of different typology (river, stream, irrigation channel) and water quality. Sampling efficiency and reliability in the calculation of E.B.I. and Quality Class by 1, 2 and 3 artificial substrates with respect to hand net have been evaluated. Influence of water quality, typology and original prevailing substrate in watercourses on the performance of artificial substrates has also been investigated. Results show a good agreement with other Authors' papers, confirming that artificial substrates represent a valid ...

Energy Technology Data Exchange (ETDEWEB)

This final report for the Swiss Federal Office of Energy looks at the possibilities for improved waste water utilization in the Lavey-les-Bains thermal spa, Switzerland. According to the regulations in force, the temperature of the waste water rejected into the Rhone river shall not exceed 30 {sup o}C, what is currently not the case. Also the operational cost shall be reduced and the waste water quality improved. The installations are presented. From the two geothermal wells, mineral water comes out at an average flow rate of 940 l/min and a temperature of 63 {sup o}C. Actual waste water data are reported. The measured thermal water consumption data, including seasonal variations, are analysed by computerized simulation and measures to reduce the consumed volume by the optimization of internal procedures are evaluated. Measures to reduce the ...

Environmental Research Database

SummaryIntended Outputs: Identification and evaluation, using participatory approaches, of demand management options:*Technical*Allocative and market based*Impact on target beneficiariesSupporting measures required when introducing water demand management options above:*Extension services & training*Water licences/rights*Institutional changes*Legal measures and regulation*Education*Support in diversifying to less water demanding activities*S [continued...]ObjectivesIdentification of the most appropriate demand management strategies for ground water abstraction, where aquifiers are being over-exploited, ensuring sustainable livelihoods of the vulnerable and poor are safeguarded. Poverty reduction strategies for areas where groundwater is being over-exploited.DescriptionProject Background: Groundwater is the principal source of both irrigation and domestic ...

British Library Electronic Table of Contents (United Kingdom)

The behavior of proton transfer facilitated by a novel thiazole derivative, N-methyl-4-(4-phenoxyphenyl)thiazol-2-amine (MPPT), across the water/1,2-dichloroethane (1,2-DCE) interface was investigated electrochemically. The ionic partition diagram for MPPT was obtained from interpretation of the cyclic voltammograms. The apparent partition coefficient of MPPT was evaluated by the shaking-flask method under experimental conditions, while that for the protonated form of MPPT was calculated from its transfer potential obtained from the ionic partition diagram. It was suggested that the mechanism for transfer of MPPT across the water/1,2-DCE) interface depends on the pH of the aqueous phase. The parameters of the facilitated proton transfer across the water/1,2-DCE interface were evaluated as ...

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 fitting method applying ...

International Nuclear Information System (INIS)

An ion chromatograph has been modified to automatically sample ten liquid water streams from the secondary side of three pressurized water reactors, Calvert Cliffs, Unit One, Rancho Seco and McGuire, Unit 1. Sampling and measurement is semicontinuous with a cycle time of approximately five hours for 10 locations with sensitivities in the range of 0.1 to 0.5 ppb. The efficiency of the condensate polishing system and subsequent transport of sodium, chloride, and sulfate around the system can be readily followed. Sulfate has been shown to have unusual volatility into the steam phase from the steam generator as well as a tendency to pass through the condensate polisher.

International Nuclear Information System (INIS)

Unit one of Daya Bay, China's first nuclear power plant was officially opened in February 1994. The nuclear island has been built by Framatome and is an improved version of the Gravelines 5 and 6, 900MWe Pressurized Water Reactors. Extra seismic protection has been included because of greater earthquake risk. The heat exchanger capacity has also been increased as the sea cooling water can be at 30"oC. The technical specifications and details of the fuel loading are given. The technical assistance, management and training of Chinese personnel are discussed. Two further units may be built if Daya Bay 1 is successful. (UK).

International Nuclear Information System (INIS)

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. (author).

Energy Technology Data Exchange (ETDEWEB)

Fuel irradiation leads to a swelling resulting from the formation of gaseous (Kr, Xe) or solid fission products which are found either in solution or as solid inclusions in the matrix. This phenomena has to be evaluated to be taken into account in fuel cladding Interaction. Fuel swelling was studied as a function of burn up by measuring the corresponding cell constant evolution by X-Ray diffraction. This study was realized on Mixed Oxide Fuels (MOX) irradiated in a Pressurized Water Reactor (PWR) at different burn-up for 3 initial Pu contents. Lattice parameter evolutions were followed as a function of burn-up for the irradiated fuel with and without an annealing thermal treatment. These experimental evolutions are compared to the theoretical evolutions calculated from the hard sphere model, using the fission product concentrations determined by the APPOLO computer code. Contribution of varying parameters influencing the ...

Energy Technology Data Exchange (ETDEWEB)

The competitiveness of a new nuclear plant vs. a new oil or gas fired combined cycle plant or a coal fired plant in East-Asia, is reviewed in the paper. Both the nuclear and the fossil fired plants are evaluated as either utility financed or independent power producer (IPP) financed. Two types of advanced light water reactors (ALWRs) are considered in this paper, namely evolutionary ALWRs (1200 MWe size) and passive ALWRs (600 MWe class). A range of capital and total generation costs for each plant type is reported here. The comparison centers on three elements of overall competitiveness: generation costs, hard currency requirements, and employment requirements. Each of these aspects is considered perspective. Year-by-Year generation cost history over the plant lifetime is shown in some cases. It is found here that a utility financed evolutionary and passive ALWRs are broadly competitive with an IPP financed gas fired ...

International Nuclear Information System (INIS)

Mixed Uranium Plutonium Carbide ((U, Pu) C), in the form of pellets encapsulated in stainless steel tubes is the fuel for Fast Breeder Test Reactor (FBTR) at Kalpakkam. For the fabrication of fuel for enlarging the core of this reactor, high burn up plutonium is used. The external exposure in these labs was significantly higher than that with low burn up Pu fuel. Dose evaluation to the organs was carried out using experimental TLDs during various operations of FBTR fuel fabrication to study the dose distribution pattern. (author)

Science.gov (United States)

Fluidized bed technology was experimentally evaluated for the combustion and gasification of cotton gin waste. The isothermal bed temperatures in the reactors could be maintained below the ash fusion point of the waste. Raw stripper harvested cotton gin trash could be metered directly into the fluidized-bed reactors indicating that little feed preparation is required. H and CO were the primary products of gasification, and approximately 3200-400 Btu of gas were produced per lb of cotton gin waste. These techniques offer the potential of providing small-scale energy conversion systems for use on farms.

Energy Technology Data Exchange (ETDEWEB)

Results indicate that the probability of double-ended guillotine break (DEGB) in the reactor coolant loop piping of Westinghouse and Combustion Engineering plants is extremely low. It is recommended that the NRC seriously consider eliminating DEGB as a design basis event for reactor coolant loop piping in Westinghouse plants. Pipe whip restraints on reactor coolant loop piping could then be excluded or removed, and the requirement to design supports to withstand asymmetric blowdown loads could be eliminated. It is also recommended that the current requirement to couple safe shutdown earthquake (SSE) and DEGB be eliminated. Recognizing however that seismically induced support failure is the weak link in the DEGB evaluation, it is recommended that the strength of component supports, currently designed for the combination of SSE plus DEGB, not be reduced. The study indicates that the probability of DEGB in ...

Energy Technology Data Exchange (ETDEWEB)

The core bypass phenomenon of borated water injected through direct vessel injection (DVI) nozzles in APR1400 (Advanced Power Reactor 1400MWe) during main steam line break (MSLB) accidents with a reactor coolant pump (RCP) running mode has been simulated using a two-channel and one-dimensional system analysis model code (MARS), and a three-dimensional computational fluid dynamics (CFD) code (FLUENT). A visualization experiment has also been performed using a scaled-down model of the APR1400. The MARS analysis has predicted a serious core bypass phenomenon of borated water, while the CFD analysis has shown results opposite to the MARS results. The CFD analysis has shown that the flow pattern in the downcomer is fully three-dimensional and that vortex flow structures are formed near the cold legs so that the borated water might pass without difficulty into the high flow region of the ...

International Nuclear Information System (INIS)

Nuclear energy provides a third of Europe's electricity with nearly no greenhouse-gas emissions. Sustained efforts are now being conducted to harmonize regulations all over Europe through WENRA and to converge on technical nuclear safety practices within the TSO network ETSON (European Technical Safety Organizations Network). In CANDU type NPP the tritiated water occurs by the neutron bombardment of deuterium. The tritiated water vapors imply health hazard (in the critical organs of the body the water presents a 10 day average biological half-life) and the early detection in nuclear plants of tritium emissions is important because the tritiated water vapors have the same characteristics as of atmospheric water vapors. By detecting tritiated vapors, the monitoring system ensures the following objectives: (a) indicates levels of tritium generally due to heavy ...

Energy Technology Data Exchange (ETDEWEB)

The first Russian reprocessing facility, known as RT-1 (located at PA-Mayak in Ural region) was started on the radiochemical plant base in 1977. Nowadays RT-1 is the sole operating reprocessing plant in Russia. The main features of RT-1 is its broad spectrum of reprocessing spent nuclear fuels (SNF). The following spent fuel types are reprocessed: -) SNF from PWR reactors (WWER-440) and FB reactor (BN-600); -) SNF of transport ship reactors; -) production reactors SNF; -) research reactor spent nuclear fuel. The world-known technological processes are used at RT-1, but there are the following distinctive features. First, the universality of the three technological lines which allows not only the reprocessing of various SNF kinds, but also to implement the combined reprocessing of some types of them. Secondly, extraction of neptunium during SNF reprocessing which is used to implement ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

Both travelers were members of a nine-person US delegation that participated in an international workshop on accelerator-based 14 MeV neutron sources for fusion materials research hosted by the University of Tokyo. Presentations made at the workshop reviewed the technology developed by the FMIT Project, advances in accelerator technology, and proposed concepts for neutron sources. One traveler then participated in the initial meeting of the IEA Working Group on High Energy, High Flux Neutron Sources in which efforts were begun to evaluate and compare proposed neutron sources; the Fourth FFTF/MOTA Experimenters' Workshop which covered planning and coordination of the US-Japan collaboration using the FFTF reactor to irradiate fusion reactor materials; and held discussions with several JAERI personnel on the US-Japan collaboration on fusion reactor materials.

International Nuclear Information System (INIS)

The ultra-high pressure waterjet technology was being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU's evaluation of efficiency and cost, this report covers the evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The ultra-high pressure waterjet technology acts as a cutting tool for the removal of surface substrates. The Husky trademark pump feeds water to a lance that directs the high pressure water at the surface to be removed. The safety and health evaluation during the testing demonstration focused on two main areas of exposure. These were dust and noise. The dust exposure was found to be minimal, which would be expected due to the wet environment inherent in the technology, but noise ...

UK PubMed Central (United Kingdom)

A second-generation solar disinfection (SODIS) system (pouch) was constructed from food-grade, commercially available packaging materials selected to fully transmit and amplify the antimicrobial properties...Full Text Available

UK PubMed Central (United Kingdom)

The procedure currently used for isolating legionellae from environmental samples recommend filtration through a 0.2-microns-pore-size polycarbonate filter. In this study we evaluated the performance...Full Text Available