Energy Technology Data Exchange (ETDEWEB)

Mitsubishi is now developing a new Pressurized water reactor (PWR), the Mitsubishi simplified PWR (MS-PWR), which has the innovative features of hybrid safety systems (an optimum combination of passive and active systems) and cooling by horizontal steam generators. In order to confirm the feasibility of the Mitsubishi hybrid safety system, various kinds of safety analyses are performed for loss-of-coolant accident events. In parallel to these safety analysis efforts, the following thermal-hydraulic tests are to be performed: (1) thermal-hydraulic test of a horizontal steam generator; (2) integrated thermal-hydraulic test using a simulation loop for the innovative MS-PWR (SLIM).

International Nuclear Information System (INIS)

Mitsubishi is developing a new type of accumulator incorporating the technology of fluidics as one of the seeds for the improved safety of the newly constructed pressurized water reactor plants. This accumulator employs a vortex flow control device, called a vortex damper, as a fluidic device to simplify the safety systems. A fundamental experimental study with a one-fifth scale model and confirmation tests with a one-third scale model to develop the vortex damper have been carried out, and satisfactory results have been achieved. The results of the confirmation tests under the prototype pressure conditions agree well with the basic tests. The flow rate ratio can be 5 to 6. The pressure loss coefficient in the large flow rate period is 8. A cavitation factor is the main parameter of the flow rate coefficient.

Energy Technology Data Exchange (ETDEWEB)

Studies technological safety of installing a water spray pressure vessel between electrostatic dedusters and coal sludge ducts. These sprays are in use elsewhere for steam generator ash removal. Dust ignition and explosion tests were carried out to examine flame and pressure wave propagation through the vessel into ducts. Water jet diameter, amount of water sprayed and coal dust removed were varied. Pressure waves exceeded 250 Pa. Test results show the vessel to be suitable for installation in briquetting plants due to its flame and explosion barrier effect and extermination of smoldering dust fires. The only disadvantage of the vessel is seen as its water and electric power consumption; about 8/sup 3//h of water and 1.5 kW/h of power per vessel serving dedusters of a 2,200 m/sup 2/ rotary brown coal dryer.

Energy Technology Data Exchange (ETDEWEB)

The scope and contents investigate and reviewed are as follows : establishment of study plan and references survey, review of related problems and inservice test standards of safety injection pump in use nuclear power plant, review of the study results in laboratory, the theoretical investigation of temperature rise according to mini-flow rate of pump, mini-flow rate working characteristics of high and low pressure injection pumps at nuclear power plants, setup of testing equipment for measuring ampere, discharge pressure and vibration, selection and behaviors analysis of major parameters concerning pump degradation.

Energy Technology Data Exchange (ETDEWEB)

A series of pipe rupture tests has been performed at JAERI to demonstrate the safety of primary coolant circuits in the event of pipe rupture in nuclear power plants. Pipe whip tests and jet discharge tests have been conducted under BWR and PWR loss-of-coolant accident (LOCA) conditions. The present paper describes the experimental and analytical results of the pipe whip tests performed under PWR LOCA conditions using 4, 6 and 8-inch test pipes. The tests were carried out at an initial pressure and temperature of 15.7 MPa and 325/sup 0/C, respectively. Moreover, a dynamic analysis of pipe whip tests was carried out using the general purpose finite element programm ADINA.

Energy Technology Data Exchange (ETDEWEB)

The authors study technological safety of installing a water spray pressure vessel between electrostatic deduster and coal sludge ducts. These sprays are in use elsewhere for steam generator ash removal. Dust ignition and explosion tests were carried out to examine flame and pressure wave propagation through the vessel into ducts. Water jet diameter, amount of water sprayed and coal dust removed were varied. Pressure waves exceeded 250 Pa. Test results show the vessel to be suitable for installation in briquetting plants due to its flame and explosion barrier effect and extermination of smoldering dust fires. The only disadvantage of the vessel is seen as its water and electric power consumption: about 8 m/sup 3//h of water and 1.5 kW/h of power per vessel serving dedusters of a 2,200 m/sup 2/ rotary brown coal dryer. (MOS).

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 ...

Energy Technology Data Exchange (ETDEWEB)

A surplus USN 40 mm ammunition can was subjected to a variety of tests. Pressure tests were carried out with nitrogen gas, followed by the venting of actual Li/SO/sub 2/ cells and batteries inside the can. A fire test was also conducted on a can packed with 10 each 10-cell batteries surrounded by vermiculite. Test results indicate the US Navy (USN) 40-mm ammunition can is suitable as a shipping container for Li/SO/sub 2/ batteries on passenger aircraft. To provide a further measure of safety, a sulfur dioxide getter was incorporated into the can. Studies indicated a commercial material, ASC carbon, is suitable for this purpose. The granular material was packaged in porous paper desiccant bags and placed in the can with the batteries and vermiculite. The batteries were vented inside the sealed can and the internal pressure monitored. ...

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled test reactor located on the Hanford Site. The FFTF is the only reactor of this type designed and operated with the intent of meeting the licensing requirements of the Nuclear Regulatory Commission (NRC). Unique characteristics of the FFTF that present special challenges related to leak rate testing include thin wall containment vessel construction, cover gas systems that penetrate containment, and a low-pressure design basis accident. The successful completion in 1986 of the third FFTF Integrated Leak Rate Test (ILRT) five days ahead of schedule and 10% under budget was a major achievement for the Westinghouse Hanford Company. The success of this operational safety test was due in large part to a special local area network (LAN) of three IBM PC/XT computers that monitored the sensor ...

International Nuclear Information System (INIS)

This standard provides requirements and guidelines for installation, inspection and testing activities that assure the quality of important mechanical parts of a nuclear power plant not covered by the ASME Boiler and Pressure Vessel Code, Section III, during construction. These parts include those mechanical systems and components whose satisfactory performance is required: for the plant to operate reliably; to prevent accidents that could cause undue risk to the health and safety of the public; or to mitigate the consequences of such accidents if they were to occur. The requirements of this standard deal with the protection and control necessary to assure that the requisite quality of those important parts of the plant are preserved from the time items are removed from storage or receiving until they are incorporated into the plant up to but not including fuel loading for PWR plants and the completion of cold functional ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Growing public and political interests towards incorporating passive safety features in nuclear installations, let Siempelkamp in late 1987 propose a solution consisting of a prestressed cast-iron pressure vessel and a passive heat removal system, integrated in the reactor cell surrounding the vessel. This solution combines the inherent safety of a prestressed metallic pressure vessel with the advantages of a passive heat removal system and thus constitutes a major step towards the goal of further reducing potential residual risks. The design had to meet the boundary conditions for reactor core and reactor building of the modular 200 MWth pebble bed reactor of Siemens/-KWU. The engineering design showed that many input parameters needed for the finite-element-analysis of the overall structure required a verification by measurements in a well scaled test setup. This was especially ...

International Nuclear Information System (INIS)

The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC

International Nuclear Information System (INIS)

The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs.

International Nuclear Information System (INIS)

This is a slide-based oral presentation given to the COG/IAEA: Fifth technical committee meeting on 'Exchange of operating experience of pressurized heavy water reactors' held in Mangalia, Romania on 7-10 September 1998. Since energization of Wolsung Unit 3 station service transformer on July 12, 1996 a line of initial test program was conducted as follows: 1. ILRT/SIT; 2. Pre-operational and Hot Functional testing with a Light Water and without Fuel in Systems; 3. Load D_2O in Moderator System; 4. Initial fuel loading; 5. Load D_2O in PHT System; 6. Hot Functional Testing with Heavy Water and Fuel in Systems; 7. Criticality and Low Power Physics Testing; 8. Power Ascension Test and, then finally, phase-D test; the plant acceptance test was accomplished after having a Mini-Overhaul to prepare for Commercial Operation. ...

International Nuclear Information System (INIS)

The Reliability Data Management System (RDMS) for safety systems of PHWR type plants has been developed and utilized in the reliability analysis of the special safety systems of Wolsong Unit 1 with plant overhaul period lengthened. The RDMS is developed for the periodic efficient reliability analysis of the safety systems of Wolsong Unit 1. In addition, this system provides the function of analyzing the effects on safety system unavailability if the test period of a test procedure changes as well as the function of optimizing the test periods of safety-related test procedures. The RDMS can be utilized in handling the requests of the regulatory institute actively with regard to the reliability validation of safety systems.

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)

Purpose: To enable safety reactor shut down upon occurrence of an abnormal excess pressure in a hydraulic control unit. Constitution: The actuation pressure for a pressure switch that generates a scram signal is set lower than the release pressure set to a pressure release valve. Thus, if the pressure of nitrogen gas in a nitrogen container increases such as upon exposure of the hydraulic control unit to a high temperature, the pressure switch is actuated at first to generate the scram signal and a scram valve is opened to supply water at high pressure to control rod drives under the driving force of the nitrogen gas at high pressure to rapidly insert the control element into the reactor and shut down it. If the pressure of the nitrogen gas still increases ...

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 criticality safety.

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Computer based control systems for safety related applications in nuclear power plants have to meet not only the functional, performance and interface requirements, they in addition, have to meet regulatory requirements like enhanced reliability, safety and security and should provide fault tolerance, diagnostics and self-supervision. The control system architecture, hardware design and software design should meet requirements as specified in design code and guides of AERB. The Dual Processor Hot Standby Process Control System (DPHS-PCS) for TAPP-3 and 4 is a safety related (Class- IB) system. DPHS-PCS regulates PHT pressure, Pressuriser pressure, Pressuriser level, Bleed condenser pressure, Bleed condenser level and Steam generator pressure. The performance, reliability and safety requirements of this control system are ...

International Nuclear Information System (INIS)

Nuclear power plant has several safety features and each safety feature is based on the operation of pumps and valves. Therefore, it is an essential basis for the safety of nuclear power plant to keep operational readiness of pumps and valves by In-Service Test (IST). According to Ministry of Education, Science and Technology (MEST) Bulletin 2008-14, the safety functioned pumps and valves of all nuclear power plants of Korea Hydro and Nuclear Power Co. Ltd.(KHNP) have been tested to verify their performance of safety function. Each safety grade pump has own design requirement and should be tested to identify whether it could meet the requirement. Design requirements and test references of all safety grade pumps of Wolsung Nuclear Power Plant Unit 2 (Wolsung ...

Energy Technology Data Exchange (ETDEWEB)

Since the integrity of the containment penetrations was confirmed under accident management (AM) conditions in the former test, failure criteria tests and aerosol trapping tests were carried out using low-voltage modules and flange gaskets of an actual plant under severe accident (SA) conditions, without AM. The safety margin for failure temperature of the penetrations and the credit for fission product (FP) aerosol trapping effect along the leakage paths of the degraded penetrations were evaluated in the present tests. Failure temperature of the penetrations ranged from 270 to 300degC for low-voltage modules and 300 to 350degC for flange gaskets under 0.4 to 1.0 MPa conditions. Pressure dependency on failure temperature was small. This means that the safety margin of failure temperature under AM condition is more than 70degC. By introducing ...

Science.gov (United States)

The hydraulic characteristics of flow control multiorifice plate assemblies designed for the FFTF reactor were investigated. The pressure drop flowrate characteristics determined in the test are presented. (JWR)

Energy Technology Data Exchange (ETDEWEB)

To improve operational safety of the fuel channel for Wolsung units the recommended primary heat transport system (PHTS) operating guidelines were developed, which provides a safe operation window in terms of pressure and temperature to maintain fuel channel integrity, which is similar to pressure-temperature curves used for primary pressure boundary integrity in PWRs and deterministic Leak-Before Break (LBB) analyses were conducted using recent understanding of delayed hydride cracking (DHC) mechanism. In addition, a comparative assessment of LBB analysis was made using the procedures of Fitness For Service Guidelines(FFSG) for pressure tubes rev. 0 and rev. 1.

International Nuclear Information System (INIS)

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) ...

International Nuclear Information System (INIS)

The possibility of developing reactor designs with inherent safety characteristics sufficient to provide ''walk away'' safety is receiving additional emphasis in the LMR program. A key element in this effort is the recognition that LMR's possess safety characteristics above and beyond those employed in past safety review processes. Some of these additional safety characteristics are due to reactivity feedback effects caused by small structural movements during hypothetical severe design transients. The effect of these characteristics upon the behavior of the FFTF under such transients has been assessed and is discussed in this paper. The paper also presents a preliminary test matrix which might allow experimental verification of the structural reactivity feedback effects. Such experimental verification should be very useful to innovative designers seeking to ...

Energy Technology Data Exchange (ETDEWEB)

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

Science.gov (United States)

... within an accuracy of +5 psig. 2. Starting Pressure and Diffuser Geometra Preliminary, tests indicated that the geometry suggested by Ref. ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. During the course of follower core assessments, TUV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core ...

Science.gov (United States)

Background Food additives are an integral part of the modern food system, but opinion polls showing most Europeans have worries about them imply an urgent need for ethical analysis of their use. Sources of data The existing literature on food ethics, safety assessment and animal testing. Areas of agreement Food additives provide certain advantages in terms of many people's lifestyles. Areas of controversy There are disagreements about the appropriate application of the precautionary principle and of the value and ethical validity of animal tests in assessing human safety. Growing points Most consumers have a poor understanding of the relative benefits and risks of additives, but concerns over food safety and animal testing remain high. Areas timely for developing research Examining the impacts of food additives on consumer sovereignty, consumer health and on ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

The 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 ...

Energy Technology Data Exchange (ETDEWEB)

The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

All the tests have been finished up in the Waste Management and Disposal Facility which has been used for the safety tests of solidified radioactive waste on sea dumping disposal. The decommissioning of this facility was performed for use of radioisotopes. This report describes the plan on decommissioning of facility for use of radioisotopes, the contamination checking methods and measurement of radioactivity, the fore case and practice for amount of generated radioactive wastes, the operation procedures for dismantlement, the safety measures, the expenses for decommissioning and so on. (author)

International Nuclear Information System (INIS)

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

Science.gov (United States)

Various methods to analyse the effect of a non-isotherme water injection on the pressure evolution during a test on a double geothermal well are investigated. Then, several types of injection test are simulated with experimental data to examine the condit...

International Nuclear Information System (INIS)

Safety-related instrument in nuclear power plants must be checked for reliability over their projected operating life. A method of conducting accelerated aging tests is presentd. It uses the Arrhenius activation energy concept and manipulation of the parameters of the test e.g. by raising test temperature, by relying on a model characterizing the chemical-related reactions of materials.

Energy Technology Data Exchange (ETDEWEB)

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only a limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri ...

Science.gov (United States)

(e.g. enclosed cooktop with built in emissions capture and treatment) . " Test hardware in laboratory, reduced pressure and analog habitat. " Objective 1. ...

Science.gov (United States)

... 37 18 Power Spectral Density Measurements of Buffet Pressures for Two Different Microphone Positions Compared with PSD Values Inferred from ...

Energy Technology Data Exchange (ETDEWEB)

The Department of Energy (DOE) policy requires that all activities be conducted in a manner that protects the safety of the public and provides a safe and healthful workplace for employees. DOE has also prescribed that all personnel be protected in any explosives operation undertaken. The level of safety provided shall be at least equivalent to that of the best industrial practice. The risk of death or serious injury shall be limited to the lowest practicable minimum. DOE and contractors shall continually review their explosives operations with the aim of achieving further refinements and improvements in safety practices and protective features. This manual describes the Department's explosive safety requirements applicable to operations involving the development, testing, handling, and processing of explosives or assemblies containing explosives. It is intended to reflect ...

International Nuclear Information System (INIS)

Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactor and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core and plant analysis was not completed in time to perform pretest ...

Energy Technology Data Exchange (ETDEWEB)

This report is the first of four volumes that identify and assess the environmental, health, and safety issues involved in using sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles that may affect the commercialization of Na/S batteries. This and the other reports on recycling, shipping, and vehicle safety are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD&D) program for Na/S battery technology. The reports review the status of Na/S battery RD&D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers cell design and engineering as the basis of safety for Na/S batteries and describes and assesses the ...

Science.gov (United States)

The synthesis of experimental data concerning the effect of jet stagnation conditions on the drag of various afterbodies is presented. Jet pressure ratio effect on boattail pressure and on boundary layer separation is analyzed. This study is made for several values of the boundary layer thickness, taking into account the fact that some test rigs make its control possible by means of tangential blowing.

Science.gov (United States)

The Interim Decay Storage tank (IDS) will be located in a concrete cell in the FFTF reactor building. The tank will have capacity to store 112 driver fuel assemblies and 10 test assemblies in sodium. A criticality safety analysis for the design of the IDS tank was performed. From the analysis, it is concluded that under normal operating conditions and minor abnormal conditions that might shift the fuel, the IDS tank will remain adequately subcritical. (auth)

Energy Technology Data Exchange (ETDEWEB)

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

UK PubMed Central (United Kingdom)

Purpose.The hypothesis that drivers with homonymous hemianopia (HH) would take a lane position that increased the safety margin on their blind side was tested with a driving simulator.Methods.Twelve...Full Text Available

Energy Technology Data Exchange (ETDEWEB)

After the completion of the 4MWth pilot plant of the Pressurizied Internally Circulating Fluidized-bed Boiler (PICFB), a total of 15 test runs, amounting to 2280 hours of operation, were concluded. The test results have confirmed the superior operating performance. Additionally, Ebara plans to develop a Pressurized Internally Circulating Fluidized-bed Gasifier (PICFG or Advanced PICFB). 8 figs.

Energy Technology Data Exchange (ETDEWEB)

This report is the first of four volumes that identify and assess the environmental, health, and safety issues involved in using sodium-sulfur (Na/S) battery technology as the energy source in electric and hybrid vehicles that may affect the commercialization of Na/S batteries. This and the other reports on recycling, shipping, and vehicle safety are intended to help the Electric and Hybrid Propulsion Division of the Office of Transportation Technologies in the US Department of Energy (DOE/EHP) determine the direction of its research, development, and demonstration (RD D) program for Na/S battery technology. The reports review the status of Na/S battery RD D and identify potential hazards and risks that may require additional research or that may affect the design and use of Na/S batteries. This volume covers cell design and engineering as the basis of safety for Na/S batteries and describes and assesses the potential ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

As part of the EPRI Motor-Operated Valve (MOV) Performance Prediction Methodology, a System Flow Model (SFM) has been developed to determine the differential pressure (DP) across MOVs as they are stroked in typical power plant systems. Input to the SFM includes valve flow characteristics (such as flow coefficient) and system characteristics (such as pump head/flow curve). The primary output is differential pressure across the valve at each stroke position. The SFM was validated by comparison with test data from the EPRI Flow Loop Test Program. In this study, we calculated DP using SFM and performed DP test for four MOVs. Compared between calculated DP and test DP of four MOV, respectively good agreements are found. (author). 6 refs., 9 figs., 1 tab.

Energy Technology Data Exchange (ETDEWEB)

In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served as the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

It is of utmost importance to have a computer code in order to analyze how different parameters (like test duration time) affect the unavailability of safety systems of nuclear. In this context, a study was performed in order to evaluate the model employed by the FRANTIC computer code, which performs detailed calculations on the contribution to the system unavailability originated by hardware failures, component tests and repairs, aiming at considering the influence of different test schemes on the system unavailability. It was shown, by means of the results attained that the numerical model used by the FRANTIC code and the analytical model proposed by APOSTOLAKIS and CHU (4) give unavailability values much similar when the component tests are supposed to be perfect. When a test is supposed to be imperfect (that is, when it may induce a test ...

Energy Technology Data Exchange (ETDEWEB)

The objective of this project was to develop a explosion suppression system capable to confine and extinguish gas explosions of the type produced in sub level caving faces when blasting to the coal pillar. Existing systems, such as triggered barriers, were considered not to be valid because of size, weight, cost, and other operational constraints. The research activities have been focused in the development of a mixed water/air spray system that should be manually activated some second before blasting. Two prototypes have been developed and tested, the first one using nozzle operating at the standard ranges of pressure that are normally available in underground coal mines, and a second one based in high-pressure nozzles. In this case, bottles containing a pressurized air/water mixtures are required. The works carried out included theoretical studies, hydraulic nozzles characterization, and modelling of ...

International Nuclear Information System (INIS)

Allowed outage time (AOT), which is required by the technical specification of nuclear power plants (NPPs), has been determined on the basis of deterministic analysis or engineering judgment. AOT is defined as the time for which safety related components can remain inoperable before a plant state is changed. Recently, plants' operating experiences and probabilistic safety assessment (PSA) results show that the AOT could be optimized. Foreign NPPs licensees have changed their technical specifications including AOT using PSA techniques. In 1998, U.S. NRC issued the regulatory guides on risk informed decision-making and technical specification changes, and these are Reg. Guide 1.174, and 1.177. The US NRC accepted AOT extension proposals including the safety injection tank (SIT) and low pressure safety injection system (LPSI) for the ABB-CE designed plants. This paper discusses interim ...

International Nuclear Information System (INIS)

The design boron concentration of the Boron Injection Tank (BIT) in Daya Bay Nuclear Power Station is 21000 #mu#g/g. The BIT should operate under high temperature to avoid boron crystallization, causing higher evaporation, frequent water makeup, higher deposition and pipe blockage to decrease the operability of the safety injection system. The author proposes to decrease the boron concentration in BIT from 21000 #mu#g/g to 7000 #mu#g/g to solve the existing problem. The safety analyses (core DNBR and containment response) are conducted and other impacts are evaluated for the BIT reduction. The analysis results show that the core DNBR meets the safety criterion and the containment pressure is within the design value for the steam line rupture accident after the BIT reduction. The feasibility study report of Daya bay BIT reduction has been approved by NNSA. The site implementation of BIT reduction has ...

International Nuclear Information System (INIS)

A program of tests were conducted in July 1986 at the Fast Flux Test Facility (FFTF) to demonstrate that the reactor could withstand a prototypic loss of flow (LOF) without scram without sustaining fuel damage. The reactor was taken to powers up to 50%, and the main primary coolant pump motors were tripped without scramming the control rods. This paper summarizes the analyses performed to demonstrate the maintenance of redundant protection for all design events as well as potential new events introduced by the test. The analyses focused on the following consequences: (1) unexpected test behavior; (2) transient overpower event during the test; and (3) LOF event during the test.

Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: A porous media solution PORFLO has been developed for the 3-dimensional two-phase flow by describing the process facility in Cartesian or cylindrical coordinates. The local porosity fraction is applied for distinguishing the fluid filled volumes from the solid structures. The solid structure contribute the two-phase flow through the wall friction, flow area and heat transfer. Optionally the solid structure may contain primary liquid of steam generators, steam in the higher temperature and pressure to be condensed or electrical heating power. By using these optional boundary conditions three different process facilities have been analysed. The thermohydraulic solution based on 5-equation approach, where the conservation equations are solved for the liquid and gas (vapour) mass, mixture momentum (giving the velocity only for the mixture), liquid and gas energy, is described shortly. In addition to that the principles modelling ...

International Nuclear Information System (INIS)

Developed under the PN II Project 71-009 - MONA, the pre-operational monitoring program of the LIL waste disposal site presented in this paper covers the main elements requested for a continuous improvement of site characterization and safety assessment, as well as for the set up of the baseline data requested for the future operational and post-operational monitoring and surveillance. These elements are: vadose and saturated zones hydrogeology, surface erosion, meteorology and radionuclides content in waters, soil and biota. Pairs of TDR (Time Domain Reflectometry) and pressure sensors, coupled with a meteorological station have been installed on site for the continuously measurement of the water content and matric potential changes due to the climate variations. The data will be used to follow the water balance, the pores water velocity and to calculate the infiltration rate. A new set of erosion rates experimentally determined by Pinhole ...

Energy Technology Data Exchange (ETDEWEB)

A 16 l liquid parahydrogen target has been developed for a measurement of the parity-violating {gamma}-asymmetry in the capture of polarized cold neutrons on protons in the {rvec n} + p {yields} d + {gamma} reaction by the NPDGamma collaboration. The target system was carefully designed to meet the stringent requirements on systematic effects for the experiment and also to satisfy hydrogen safety requirements. The target was designed to preserve the neutron polarization during neutron scattering on liquid hydrogen (LH{sub 2}), optimize the statistical sensitivity to the {rvec n} + p {yields} d + {gamma} reaction, minimize backgrounds coming from neutron interaction with the beam windows of the target cryostat, minimize LH{sub 2} density fluctuations which can introduce extra noise in the gamma asymmetry signal, and control systematic effects. The target incorporates two mechanical refrigerators, two ortho-para convertors, an aluminum cryostat, an aluminum target ...

International Nuclear Information System (INIS)

The ISI (In-Service Inspection) system has been developed so as to maintain the structural integrity of heat transfer tubes in the primary pressurized water cooler in the HTTR (High Temperature Engineering Test Reactor). This system consists of eddy current probes, ultra-sonic probes, insertion and extraction units, positioning unit and so on. Verification and performance tests of the developed ISI system were carried out using mock-up heat transfer tubes in the primary pressurized water cooler. The constitution of the system, R and D results of the inspection probes, and verification and performance test results of the ISI system for heat transfer tubes are described in this paper. (author)

Energy Technology Data Exchange (ETDEWEB)

The public expects that the Los Alamos National Laboratory (LANL) will operate in a manner that prevents negative impacts to the environment and protects the safety and health of its employees and the public. To achieve this goal within budget, the Department of Energy (DOE) and LANL must develop new and improved environment, safety, and health (ES and H) technologies and implement innovative, more cost-effective ES and H approaches to operations. In FY95, the Environment, Safety, and Health (ESH) Division initiated a Technology Development, Evaluation, and Application (TDEA) program. The purpose of this unique program is to test and develop technologies that solve LANL ES and H problems and improve the safety of LANL operations. This progress report presents the results of 10 projects funded in FY97 by the TDEA Committee of the Environment, Safety, and Health ...

Energy Technology Data Exchange (ETDEWEB)

As a basic research to consider aging deterioration of the operating nuclear power plant to seismic fragility analysis, aging deteriorations occurring safety related equipment of both Kori unit 1 and Wolsung unit 1, are investigated in this study. First of all, 378 and 152 safety related equipment are selected at Kori unit 1 and Wolsung unit 1 respectively. Seismic review team including seismic capability engineer, is organized and seismic walkdown is carried out using the nondestructive tests. As a results of seismic walkdown, crack is a typical aging deterioration which can reduce the seismic safety of safety related equipment and the other aging deteriorations such as concrete compressive strength, corrosion, and tightness of anchor bolt, have a much smaller influence than crack. In order to manage the aging deterioration data collected through the seismic walkdown in effective ...

Science.gov (United States)

A schematic of the Bench cavity receiver system is given, and the primary objectives, schedule and research plan for each of four tests are discussed. The tests are: (1) cold flow testing to functionally check all valves and their response using high pressure nitrogen gas; (2) a hot checkout to simulate both steady state and transient flux inputs to the receiver which are expected at the French facility; (3) basic performance using solar energy at the French facility; and (4) ir radiant heat testing. (LEW)

Science.gov (United States)

In order to reduce the potential for severe liquid pipeline accidents, the U.S. Materials Transportation Bureau (MTB) proposes to require a hydrostatic test on all onshore pipelines carrying highly volatile liquids which have not been previously tested to at least 1.25 times their maximum operating pressure for at least 24 hr. Comments should be received by the MTB by 2/15/79. Late filed comments will be considered as far as practicable.

International Nuclear Information System (INIS)

In the VEGA program on radionuclide release from irradiated fuel under severe accident conditions, 10 tests in total were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed the increase in release rate above 2,800 K or at the fuel liquefaction and the decrease in release rate under elevated pressure, which was a first observation in the world. The data on low-volatility radionuclide release, release from MOX fuel, effect of fuel oxidation, and eutectic reaction with cladding on release were obtained from the tests. The mechanism of pressure effect on release was examined and a new release model with pressure effect was proposed. In addition, the pressure effect on source term evaluation and effectiveness of accident management ...

International Nuclear Information System (INIS)

The FFTF is a 400-MWt sodium-cooled fast neutron flux test reactor located on the US government-owned Hanford Site in southeastern Washington state. The reactor is operated for the US Department of Energy by the Westinghouse Hanford Company. Since FFTF started routine operation in 1982, the commercially fabricated driver fuel has performed flawlessly to well beyond the design goal peak burnup of 80 MWd/kgM. The core average discharge exposure is now some 60% beyond the original design expectations and attests to the ruggedness and reliability of the mixed oxide fuel system. In Cycle 9 sixteen long-life assemblies were installed to begin the irradiation of mixed oxides in the advanced low-swelling alloy HT-9 as the Core Demonstration Experiment (CDE). Operation of the plant from initial startup testing to ten cycles of operation has confirmed that the nuclear characteristics are well within the design predictions, and all parameters have ...

Energy Technology Data Exchange (ETDEWEB)

A back-up system is provided for a pumping plant or a pressure reservoir system used to maintain insulating oil under pressure in high voltage submarine power cables which includes a gas driven membrane pump to insure a small outflow of oil at a rupture point in the power cable to prevent water from penetrating the cable. The proper operation of the membrane pump is insured by introducing degasified oil at the drive side of the pump during the stand-by and test modes of operation of the back-up system. 11 drawings.

Science.gov (United States)

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

Energy Technology Data Exchange (ETDEWEB)

The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix.

British Library Electronic Table of Contents (United Kingdom)

Biomarkers are biometric measurements that provide critical quantitative information about the biological condition of the animal or individual being tested. In drug safety studies, established toxicity biomarkers are used along with other conventional study data to determine dose-limiting organ toxicity, and to define species sensitivity for new chemical entities intended for possible use as human medicines. A continuing goal of drug safety scientists in the pharmaceutical industry is to discover and develop better trans-species biomarkers that can be used to determine target organ toxicities for preclinical species in short-term studies at dose levels that are some multiple of the intended human dose and again later in full development for monitoring clinical trials at lower therapeutic ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The meeting featured seven sessions with 18 papers abstracted/indexed therein as follows: keynote address: tailings dams safety - implications for the dam safety community; 1 - design and performance: performance monitoring of dams: are we doing what we should be doing?; tailings dams from the perspective of conventional dam engineering; and design overview of Syncrude's Mildred Lake east toe berm; 2 - design and modelling: use of a 2D model for a dam break study on the ALCAN hydroelectric complex in Quebec; and spillway design implications resulting from changes in rainfall extremes; 3 - risk and dam safety I: closing the gaps in the dam safety guidelines; the reality of life safety consequence classification; and surveillance practices for the next millenium; 4 - risk and dam safety II: quantitative risk-assessment using the capacity-demand analysis; ...

Energy Technology Data Exchange (ETDEWEB)

The Control Rod Drive(CRD) nozzles for PWR nuclear power plants(NPP) house the control rod drives. The number of nozzle penetrations range from the mid-30's to over 100 in each reactor head. The integrity of CRD nozzles is very important, because the primary pressure boundary is established with the J-groove weld joining the nozzle to the head clad surface. The Alloy 600 PWSC CRD nozzle leaks discovered in the fall of 2000 and spring of 2001 in several US plants. Therefore the NRC has recommended a more proactive effort by US utilities to inspect similarly susceptible nozzles in all US plants. The primary safety concern is circumferential cracks that can permit the nozzles to separate from the head at high velocity and produce a large-break leak in the reactor vessel. A secondary concern is head leakage from any through-wall cracks in the nozzle or J-groove weld area. Numerous inspection and repair tools have been developed to address ...

Energy Technology Data Exchange (ETDEWEB)

Since the first commercial operation in 1983, Wolsung unit 1 has experienced several aging phenomena, especially in pressure tubes and feeder pipes. In case of pressure tubes, in-service inspections (ISI) revealed that a major portion of inspected tubes was in contact with calandria tubes. The likelihood of blister formation was a safety concern because it is a potential threat to pressure tube integrity. Wolsung unit 1 was, therefore, subjected to the SLAR (Spacer Location And Reposition) work to separate the contacted pressure tubes from calandria tubes. In this paper, experience of in-service inspection to the pressure tubes will be discussed including the irradiation creep elongation and CIGAR (Channel Inspection and Gauging Apparatus for Reactors) examination of pressure tubes. On the other hand, the problem of wall thinning in the ...

Science.gov (United States)

Test firings of 120-mm rounds were performed to characterize effects due to the combustible cartridge cases used. Comparisons were made between measured ballistic parameters, especially gas pressures, and those predicted by computerized simulation calcula...

Energy Technology Data Exchange (ETDEWEB)

The author describes the history of the Soviet Unions nuclear weapons testing in Siberia from the first bomb on 29th August 1949 until the test-ban treaty of 1963. The effects of the weapons testing on the local population is only now becoming widely known. Levels of cancers and cataracts are significantly higher than in uneffected populations, and there are also high levels of some lung, bone, skin and digestive conditions. Perhaps the most important information relates to cataracts, previously thought to be caused by single large bursts of radiation. Regulatory authorities worldwide will need to tighten the safety limits for eye radiation exposure to these findings. (U.K.).

International Nuclear Information System (INIS)

A vortex diode has been developed as a reverse flow limiter in the primary circuit of an advanced gas cooled reactor. In addition to the development work on a prototype diode to optimise performance and geometry, measurements were also made on an available experimental diode of similar size with pressure differences up to 4 MPa and temperatures up to 600 K using nitrogen, argon and carbon dioxide as the test fluids. Correlation of data from all tests was satisfactorily obtained using isentropic one-dimensional nozzle flow equations. (author).

Energy Technology Data Exchange (ETDEWEB)

Automatic ultrasonic testing is presented as a method for recurrent inspections of pressure gas vessels, from validation to practical testing to customised test applications. National and international regulations and standards are presented along with related implementation problems. [German] Es wird die automatisierte Ultraschallpruefung im Bereich der wiederkehrenden Pruefung von Druckgasbehaeltern, als Alternative zur herkoemmlichen, klassischen Wasserdruckpruefung, vorgestellt. Von der Validierung ueber die praktische Erprobung bis hin zur kundenspezifisch entwickelten Pruefapplikation. Es werden die Anforderungen aus nationalen und internationalen Vorschriften und Normen sowie die damit zusammenhaengenden Umsetzungsprobleme behandelt.

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Containment is considered as ultimate safety system and is designed to enclose whole reactor system and prevent the spread of active air-borne fission products. For Pressure and Temperature calculation, Design Basis Accident (Dba) is double ended break of reactor inlet header or main steam line break but activity release studies are done to access its performance following limiting design basis accident i.e. Loss of Coolant Accident (LOCA) and Emergency Core Cooling System (ECCS). In such accident scenario, the core is severely damaged and results in production of steam and hydrogen along with release of activity to containment environment. Containment functions are maintained in such accident, and radiological consequences are within the prescribed limits. (author)

Science.gov (United States)

This double-blind, multicenter, randomized placebo-controlled study evaluated the antihypertensive efficacy and safety of nebivolol, a selective beta1-adrenoreceptor blocker with vasodilating effects, in patients with mild to moderate hypertension (sitting diastolic blood pressure [SiDBP] > or =95 mm Hg and < or =109 mm Hg). A total of 909 patients were randomized to receive placebo or nebivolol 1.25, 2.5, 5, 10, 20, or 40 mg once daily for up to 84 days. The primary end point was the change in trough SiDBP from baseline to study end. Nebivolol significantly reduced trough SiDBP (8.0-11.2 mm Hg compared with 2.9 mm Hg with placebo; P<.001) and trough sitting systolic blood pressure (a 4.4-9.5-mm Hg decrease compared with a 2.2-mm Hg increase [corrected] with placebo; P< or =.002). The overall adverse event experience was similar in the nebivolol (46.1%) and placebo (40.7%) groups (P=.273). Once-daily nebivolol ...

Energy Technology Data Exchange (ETDEWEB)

One of the more extensive uses of directional drilling anywhere in the world has been in the development of the East Wilmington Oil Field in Long Beach, California. The Measurement-While-Drilling (MWD) directional system has been tested on several wells and proven to provide the required accuracy, along with many advantages over past methods, used in the field development. Accurate transmission by MWD of bottom-hole measurements to the surface is provided by mud pressure pulses generated in the drill pipe downhole and detected by a pressure transducer includes the means for detecting, recording and processing these pressure pulses, to translate the information from the pressure pulses to rig floor displays usable by the drilling crew. 2 refs.

International Nuclear Information System (INIS)

Corrosion of steam generator heat transfer tubes (SGHTT) is one of the important problems which affect safety operation of nuclear power plants (NPP), and the hazard of pitting corrosion of heat transfer tubes is the most serious. With an acoustic emission device, the signals during a corrosion test on SGHTT were collected and analyzed, and the corrosion points in the tubes were located accurately. The results show that pitting corrosion of heat transfer tubes has passed through three periods in its development: expansion phase, stationary phase and rapid developing phase. The corrosion damage of HTT can be found earlier with acoustic emission than any other non-destructive testing methods. Acoustic emission can be used for on-line and real-time monitoring of the safety and operation of the steam generator and has therefore a great significance. (orig.)

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a liquid-metal-cooled fast reactor, which is designed to test a variety of different structural and fuel materials. A safety analysis is performed for each experiment that is irradiated in FFTF. The FFTF final safety analysis report (FSAR) assumed that all driver fuel assemblies would maintain cladding integrity during normal operations and all design transients. Maintenance of cladding integrity retains three barriers to any fission gas release to the public and also prevents any potential contact between the fuel and coolant. Experiments are, in general, expected to meet the same criterion. Selected experiments can, however, be classified as run-to-cladding-breach experiments (RTCB). The purpose of this paper is to describe alternative acceptance criteria for RTCB experiments that they feel provide protection equivalent to the maintenance of cladding integrity.

International Nuclear Information System (INIS)

Emergency Core Cooling System (ECCS) is a safety critical system provided to mitigate the consequence of Loss of Coolant Accident (LOCA) in PHWR. Unlike 220MWe, all header injection has been introduced in 540MWe to simplify the logic. ECCS Integrated Test is schematic approach to establish that ECC system will behave as per design intent during actual LOCA condition. Objective of ECCS Integrated test is to ascertain that various ECC system components operate as intended in design. Additionally, the various system resistances which form the input to LOCA analysis are validated. This test has been carried out by creating actual LOCA during cold and pressurised condition of PHT system to establish all phases of injection with overlap. This paper discusses the results obtained during the Integrated Test and comparison with the prediction during the commissioning of first unit of 540 ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

The recent definition of a postulated thermal shock accident followed promptly by system repressurization, termed an overcooling or pressurized thermal shock accident, has set a large analysis and research effort into motion. The essential elements are concerned with defining the accident transients, evaluating the instrumentation and controls that cause the postulated accidents, and evaluating the metallurgical and structural mechanics aspects of the reactor vessel with respect to its failure potential. This paper poses the question faced by the Nuclear Regulatory Commission (NRC) for the vessel steel embrittlement, annealing, and surveillance dosimetry facets of this postulated accident and provides information on our plans for study of this problem as well as current status.

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

Some recent achievements in the field of high pressure vessels and safety devices have offered a concrete chance to the application of compressed hydrogen for fleet commercial vehicles for urban use. Accordingly with this concept, ENEA has modified a Fiat Ducato van with a dual fuel engine, retaining the gasoline tank for long distance travelling and adopting the external mixture formation technique, with hydrogen injectors developed by ENEA, for non-polluting short-range duties in urban traffic. The article deals with the rationale for this choice and gives a general view of the project. (author)

Energy Technology Data Exchange (ETDEWEB)

The properties of polycrystalline (Ti, Al)N coatings deposited on non-nitrided, classically plasma-nitrided and low pressure plasma-nitrided AISI H11 steel samples were investigated. The plasma deposition and low pressure plasma nitriding were performed in a Z700-LH magnetron sputter ion plating unit, while a separate unit was used for plasma nitriding of specimens at a pressure of several millibars. The (Ti, Al)N coating was deposited onto all the samples using the same equipment as for the plasma deposition and low pressure plasma nitriding. For the characterization of the composite structures, the following methods were used: scratch test, X-ray diffraction analysis, scanning electron microscopy, scanning tunnelling microscopy and microhardness testing. It was found that plasma nitriding prior to coating deposition strongly affects the growth and properties ...

Energy Technology Data Exchange (ETDEWEB)

The Institute of Gas Technology (IGT) has conducted a fundamental research program to determine the technical feasibility of a process concept that involves pressurized CO{sub 2}-coal gasification followed by a high-temperature CO{sub 2} removal step, using a MgO-based sorbent. To investigate a hot-CO{sub 2}-removal process, various MgO-containing materials were tested for their reactivity and capacity for removing CO{sub 2} from hot gas streams in a high-pressure thermobalance. No naturally-occurring materials were found to be acceptable, but a KOH-promoted magnesia was found to be sufficiently active for CO{sub 2} removal and to have a good capacity for CO{sub 2} absorption. Tests were conducted at pressures and temperatures in the ranges of 100--300 psig and 850{degree}--1170{degree}F. The CO{sub 2} absorption/desorption cycle was shown to be achievable by a simple ...

International Nuclear Information System (INIS)

Full text: Korea Atomic Energy Research Institute (KAERI) has many R and D facilities in operation. including HANARO research reactor, radioactive waste treatment facility (RWTF), post-irradiation examination facility (PIEF) and irradiated material test facility (IMEF). Recently. nation-wide interest is focused on the safety and security of major industrial facilities. Safe operation of nuclear facilities is imperative because of the consequence of public disaster by radiological release/contamination, in case of an accident. Recently, Ministry of Science and Technology (MOST) of the Korean government announced amendments of Atomic Energy laws to enforce requirements of the physical protection and radiological emergency. All provisions on nuclear safety regulation and radiation protection are entrusted to the Atomic Energy Act(AEA). The Act is enacted as the main law concerning the safety regulation of ...

Energy Technology Data Exchange (ETDEWEB)

In the atmospheric pressure coal combustion area, the Internally Circulating Fluidized Bed Boiler (ICFB) had already been proven to possess superior performance features in terms of the ease with which load conditions can be controlled and because of its low pollution level. Based on the particular merits of the ICFB furnace and the application of this technology to pressurized fluidized bed boiler technology, it was possible to achieve a pressurized fluidized bed boiler with superior ease of operation, reliability and low-pollution characteristics. From fiscal 1992, the authors have conducted extensive research work, including cold model tests and computer simulations to investigate the possibility of applying ICFB technology to the pressurized fluidized bed boiler technology in an attempt to develop the Pressurized Internally Circulating Fluidized Bed Boiler ...

International Nuclear Information System (INIS)

During the operation of the Fast Flux Test Facility (FFTF), radioactive gases, primarily xenon and krypton, will be produced which will require processing and storing. Two systems have been installed in the FFTF for handling these gases: (1) one to handle, primarily, the reactor cover gas system, and (2) a second to handle the cells and cover gas systems, other than the reactor, whose atmosphere may become contaminated. The system that processes the reactor cover gas, which is argon, is called the Radioactive Argon Processing System (RAPS). The effluent argon from RAPS will normally be sufficiently decontaminated to allow its reuse as the reactor cover gas. If the radioactive level in the RAPS becomes too high, the exhaust stream will be diverted to the Cell Atmosphere Processing System (CAPS), a system which can function as a backup to RAPS. The design and operation of the RAPS and CAPS systems are described and certain safety aspects of the ...

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Pipe whip tests or jet discharge tests have been performed at the Japan Atomic Energy Research Institute, which simulate the instantaneous guillotine break of primary coolant piping in nuclear power plants. This paper describes the results of the 4-inch pipe whip tests(RUN 5407, 5501, 5504, 5603), under the BWR LOCA conditions, which were performed from 1979 to 1981. The test pressure was 6.8 MPa and test temperature 285/sup 0/C. In these tests, clearance was kept constant at the value of 100 mm and overhang lengths were 250, 400, 550 and 1,000 mm, respectively. The main purpose of these tests is to investigate the effect of overhang length on pipe whip behavior. From the tests results, the pipe movement is effectively limited by the restraints if the overhang length is 250 mm or 400 mm. The ...

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The safe function of a new pipe whip restraint device has been demonstrated in a full scale test. The restraint is based on using a shape memory alloy to protect a pipe and its environment in the event of a double-ended-guillotine-break. The evaluation test has been performed at boiling water reactor (BWR) operating pressure and temperature using a pipe representing BWR primary piping. (orig.) 2 refs.

International Nuclear Information System (INIS)

In-service Test (IST) of the nuclear power plant is very important to maintain the safety of the plants. The safety features of nuclear power plant are based on the operation of pumps and valves. Therefore, it is an essential basis for the safety of nuclear power plant to keep operational readiness of pumps and valves. Because of the importance of the IST, most of nuclear power plants should designate the IST department and personnel. Because the nuclear power plants of Korea have several types and the designs and operations are different from each other, the methods and experiences are various for the management of the IST program. However, there are no medium to communicate with each other and exchange the tips and information about IST program and education and discussion would be required to apply a consistent code. The nature of the IST Standardization Management Data Base (SMDB) is to exchange ...

International Nuclear Information System (INIS)

In the early sixties it was decided to investigate salt formations on its suitability to host heat generating radioactive waste in Germany. In the reference repository concept consequently the emplacement of vitrified waste canisters in deep vertical boreholes inside a salt mine was considered whereas spent fuel should be disposed of in self shielding casks (type POLLUX) in horizontal drifts. The POLLUX casks, 65 t heavy carbon steel casks, will be laid down on the floor of a horizontal drift in one of the disposal zones to be constructed in the salt dome at the 870 m level. The space between casks and drift walls will be backfilled with crushed salt. The transport, the handling und the emplacement of POLLUX casks were subject of successfully performed demonstration and in situ tests in the nineties and resulted in an adjustment of the atomic law. The borehole disposal concept comprises the emplacement of unshielded canisters with vitrified HLW in boreholes with a ...

International Nuclear Information System (INIS)

This paper demonstrates that a commercial-scale facility for encapsulating krypton-85 in zeolite-5A or glass at a 2000 MTHM per year nuclear fuel reprocessing plant can be designed to contain fragments and the 340 to 850 kCi krypton-85 inventory from an assumed catastrophic failure of the high pressure vessel. The vessel failure was assumed as a worst case and was not based on a detailed design evaluation or operating experience. The process design is based on existing commercial hot isostatic pressing technology operated at up to 40 times the scale required for krypton encapsulation. From the calculated process gas inventory in the pressure vessel and vessel design, the maximum explosive energy of 8.4 kg TNT and resulting vessel plug and fragment velocities were calculated. The facility Containment Cell housing the high pressure vessel was designed to contain the gases, fragments, and the shock wave energy calculated for a ...

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A low pressure nuclear thermal rocket (LPNTR) is configured to meet the requirements of a nuclear stage for manned Mars exploration. Safety, reliability and performance are given equal consideration in selecting the stage configuration. Preliminary trade studies are conducted to size the engine thrust and determine the thrust chamber pressure. A weight breakdown and mechanical configuration for the selected LPNTR concept are defined. A seven engine stage configuration is selected which gives a two engine out capability and eliminates the need for engine gimbaling. The stage can be ground assembled and launched as a unit including tankage for trans Earth injection and Earth orbital capture. The tankage is configured to eliminate the need for an inert shield. The small engine will be cheaper to develop than a single engine providing full thrust, and will be compatible with stages for Earth orbital, Lunar and deep space ...

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No information on steam/water two-phase flow behavior in large diameter pipes (10 inch or larger) at elevated pressures is available in the open literature. However, there are many applications, in the nuclear, chemical and petroleum industries among others where two-phase flows in large diameter pipes at elevated pressures and temperatures are encountered routinely or under accident scenarios. Experimental data on steam-water two-phase flow in a large diameter (20 inch, 50.08 cm I.D.) vertical pipe at elevated pressures and temperatures (2.8 MPa/230 C--6.4 MPa/280 C) have been obtained. Void fraction, two-phase mass flux, phase and velocity distributions as well as pressure drop along the test pipe have been measured using the Ontario Hydro Technologies (OHT) Pump Test Loop. The void fraction distributions were found to be axially symmetric and nearly flat over ...

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A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pin are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

International Nuclear Information System (INIS)

The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year operational performance of the FFTF reactor is discussed in this report. 6 refs., 10 ...

International Nuclear Information System (INIS)

This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80"+trademark Standard Design. This volume 11 discusses Radiation Protection, Conduct of Operations, and the Initial Test Program.

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In a slurry explosive, the fine droplets of the aqueous phase are dispersed in a continuous oily phase which comprises a thin film surrounding each droplet, the latter being less than one micron in size. The authors present these commercial explosives in three different forms, indicate their explosive properties and discuss their advantages from the safety viewpoint. They also report on tests undertaken with slurry explosives in quarries possessing different types of rock and using deep vertical shotholes.

International Nuclear Information System (INIS)

Any transport of nuclear material is associated with the risk of contamination after release into working areas or environment. stationary installed safe geometry vessels with pneumatic transfer between them offer unique safety features and reduce operating costs. The article describes the case of HTR fuel spheres, where a specially designed conveying system has been developed and the prototype conveyor has been tested.

International Nuclear Information System (INIS)

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

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There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post-Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And ultrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughly because of narrow and confined accessibility, that is , an inspection space between the pressure tube channels is less than 100 mm and pipes in ...

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A novel device called the gas expansion model (GEM) is being developed at the Hanford Engineering Development Laboratory for testing in the 400-MW(th) fast flux test facility (FFTF) reactor. Incorporation of the GEM into liquid-metal reactor designs is intended to measurably contribute to the achievement of inherent safety, by allowing the reactor to passively shut down even in the extremely remote (hypothetical) event of an unprotected (no scram) loss-of-flow accident. The purpose of this paper is to describe the GEM and present predictive analyses of the effectiveness of the device during unprotected loss-of-flow experiments in the FFTF.

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This report presents the results of analysis of selected data from the NRU test series dealing with heatup and reflood heat transfer during postulated PWR LOCA conditions. These tests used nuclear fuel rods and some considered clad ballooning and rupture. Also included was an electrically-heated rod ballooning test, REBEKA-6. The COBRA-TF computer program, renamed PYTHONS, was modified and used for the analytical tool. Modifications included provisions for fuel rod gas flow and pressure, creep deformation and rupture, channel blockage, and blockage heat transfer. Calculated clad temperatures for NRU unpressurized rods show quite good agreement with experimental data. The calculated amount and axial extent of clad ballooning for pressurized rods agrees reasonably well with post-test examinations of the NRU bundles. Time to failure was underpredicted in the MT-3 ...

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General-Purpose Heat Source radioisotope thermoelectric generators (GPHS-RTGs) will provide electrical power for the NASA Galileo and European Space Agency (ESA) Ulysses missions. Each GPHS-RTG comprises two major components: GPHS modules, which provide thermal energy, and a thermoelectric converter, which converts the thermal energy into electrical power. Each of the 18 GPHS modules in a GPHS-RTG contains four /sup 238/PuO/sub 2/-fueled capsules. LANL conducted a series of safety verification tests on the GPHS-RTG before the scheduled May 1986 launch of the Galileo spacecraft to assess the ability of the GPHS modules to contain plutonia in potential accident environments. As a result of the Challenger 51-L accident in January 1986, NASA postponed the launch of Galileo; the spacecraft launch vehicle was reconfigured and the spacecraft trajectory modified. These actions prompted NASA to reevaluate potential mission accidents and the extended ...

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

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Many of today's automated real-time testing systems for embedded systems were developed using expensive custom hardware and software. In this article they describe how to use commercially available off-the-shelf hardware and software to design and develop an automated real-time test systems for Embedded Programmable Logic Controller (PLC) Based Control Systems. The system development began with the implementation of the VALI/TEST Pro testing methodology as a means for structuring the testing. Using this methodology, they were able to decompose system requirement documents for a Personnel Safety System (PSS) into its high, intermediate and detail level requirements. next, the validation procedures for the PSS system were decomposed into testing units called builds, test runs and test cases. ...

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For a better understanding of sulphidation mechanisms, some alloys (alloy 600, alloy 800, Uranus S, iron-chromium-aluminium-alloy) and metals (iron, chromium, nickel, molybdenum, titanium) were tested at 773, 873, 1 073 K in gaseous sulphur dioxide. Total pressure was 760 torrs. Sulphur dioxide pressure was 760, 100 and 10 torrs. Argon, oxygen, water vapor were used, successively, as pressure complement. Oxygen supply, generally speaking cause decay of corrosion resistance in SO/sub 2/ atmospheres especially at 1 073 K with low ratio p/sub O2//P/sub SO/sub 2//. Water vapor supply act similarly. Temperature laws and pressure laws do not give monotonous rise of corrosion values. There are maximas suggesting corrosion products undertaking a volatilization process. According to test conditions, Uranus S, chromium, alloy 800 (and titanium at some extent) showed good ...

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A characteristic trait of Francis turbines operating at low-head is pressure pulsations that occur during certain load levels of the generator. These stem from pressure variations across the turbine due to pulsating flow in the draft-tube. This surging action of the water column is related to draft-tube geometry and flow-rate of water in the penstock. The pressure pulsations cause torque variations on the turbine and corresponding electric power pulsations. If electric power is used as a feedback signal to the Power System Stabilizer (PSS), then Mvar and terminal voltage pulsations will occur when the generator is operating in the ``rough zone``. This paper describes field test results for investigating feedforward control from the penstock, draft tube and spiral case pressure to reduce the effects of Mw pulsations on PSS output signals. This investigation involved a PSS with ...

International Nuclear Information System (INIS)

The alloy type 'alloy 800' with its different variants is used for numerous components in high-temperature reactor (HTR) plants, e.g. for hot-gas conductors, the high-pressure part of steam generators, superheaters, steam collecting vessels and live steam conductors. This study deals with the evaluation of creep rupture tests on 28 welded joints of alloy 800 with an accumulated test duration of 410 years, with the objective to obtain quantitative information on a possible loss of strength in comparison with the unwelded material. The longest test period of an individual test amounted to 98,000 h and the analyzed test temperature range lay between 550"0 and 1000"0C. (orig./MM).

International Nuclear Information System (INIS)

Extensive experimental and analytical investigations of fluid flow and heat transfer in gas-cooled rod bundles have been carried out. Different bundle geometries with partially or fully roughened rod surfaces were tested in a carbon dioxide loop. An advanced and comprehensive measuring control and instrumentation are important design features of this experiment. Comprehensive thermal hydraulic subchannel analysis computer codes have been developed in order to assist fuel element design calculation for gas-cooled reactors. The experiments, codes and their verification procedure are described and the results of comparisons between measured and calculated pressure and temperature distributions are given. (orig.).

International Nuclear Information System (INIS)

Series of pipe rupture tests have been performed at the Japan Atomic Energy Research Institute to demonstrate the safety of the primary coolant circuits in the event of pipe rupture in nuclear power plants. The pipe whip tests have been conducted to study the dynamic response of the pipe and restraints. The results of the pipe whip tests using test pipes of 4-inch in diameter under the BWR LOCA conditions (285"0C, 6.8 MPa) were reported in the previous paper F8/5 of the 6th SMiRT. The present paper describes the results of the pipe whip tests using test pipes of 6-inch in diameter. The test pipe was made of Type 304 stainless steel and was 165.2 mm in outer diameter and 11.0 mm in thickness, and was fixed at the pipe support so that the length of the test section was 3000 mm. The restraints were made ...

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The experimental technique used for detection of subcooled boiling through analysis of the fluctuation contained in pressure transducer signals is presented. This work was partly conducted at the Institut fuer Kerntechnik und zertoerungsfreie Pruefverfahren von Hannover (IKPH, Germany) in a thermal-hydraulic circuit with one electrically heated rod with annular geometry test section. Piezo resistive pressure sensors are used for onset of nucleate boiling (ONB) and onset of fully developed boiling (OFDB) detection using spectral analysis/ signal correlation techniques. Experimental results are interpreted by phenomenological analysis of these two points and compared with existing correlation. The results allow to conclude that this technique is adequate for the detection and monitoring of the ONB and OFDB. (author)

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A detailed chemical kinetic model to describe the autoignition of gasoline surrogate fuels is presented consisting of the fuels iso-octane, n-heptane, toluene, diisobutylene and ethanol. Model predictions have been compared with shock tube ignition delay time data for surrogates of gasoline over practical ranges of temperature and pressure, and the model has been found to be sensitive to both changes in temperature and pressure. Moreover, the model can qualitatively predict the observed synergistic and antagonistic non-linear blending behaviour in motor octane number (MON) for different combinations of primary reference fuels (PRFs) and non-PRFs by correlating calculated autoignition delay times from peak pressures and temperatures in the MON test to experimental MON values. The reasons for the blending behaviour are interpreted in terms autoignition chemistry. 37 refs., 11 figs., 4 tabs.

International Nuclear Information System (INIS)

This paper discusses how the U.S. Liquid Metal Reactor Development Program has been restructured to carry out R and D on advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the integral fast reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has been made in demonstrating both the metallic fuel and the new electrochemical processes of the IFR. The FFTF facility is converting to metallic fuel; however, FFTF also maintains a considerable U.S. program in oxide fuels. In addition, generic programs are continuing in steam generator testing, materials development, and with international cooperation, aqueous reprocessing.

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

A large portion of the work performed at nuclear power plants follows standardized flow paths. For example, although components on which the maintenance crew works differ from one assignment to the next, all assignments basically follow the same process: requesting, reviewing, planning, scheduling, executing, testing, and documenting the work. In general, the term open-quotes work processclose quotes is used to refer to a standardized sequence of tasks designed within the operational environment of an organization to achieve a specific goal. The predictable nature of work processes suggests that a systematic analysis can be conducted to identify the desirable design of the process and to develop performance measures with respect to the strengths and weaknesses in the process. Furthermore, because of the close relationship of the work process to plant performance and plant safety, it is believed that such an analysis will facilitate the ...

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A hypothesis on the transition of a fully water-saturated coal bed methane reservoir from single-phase flow through unsaturated flow to the two-phase flow regime was developed and substantiated using field data. The concept of parameter measurement windows was introduced to obtain the required reservoir parameters from flow testing. A new empirical relationship for capillary pressure vs. saturation, referred to as the Fermi distribution, was developed and shown to fit the capillary pressure-saturation data for coal, sandstone, and other consolidated and unconsolidated materials. New approximate analytic solutions for flow to a sink from an infinite coal bed reservoir with desorbing methane have been developed and verified using numerical models for the following cases: (1) single-phase gas flow with desorption; (2) single-phase water flow in fully water-saturated and unsaturated flow regimes. A numerical model based on ...

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Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under ...

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 dose C) Seismic isolation technologies to standardize plant design independent from site conditions D) ...

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Lubricant additives have been known to affect rolling element bearing surface durability for many years. Tapered roller bearings were used in fatigue testing of lubricants formulated with gear oil type additive systems. These systems have sulfur- and phosphoruscontaining compounds used for gear protection as well as bearing lubrication. Several variations of a commercially available base additive formulation were tested having modified sulfur components. The variations represent a range of ''active'' extreme pressure (EP) chemistries. The bearing fatigue test results were compared with respect to EP formulation and test conditions. Inner ring near-surface material in selected test bearings was evaluated on two scales: the micrometer scale using optical metallography and the nanometer scale using transmission electron ...

International Nuclear Information System (INIS)

Development in Instrumentation and Control (I and C) technology provides more convenience and better performance, thus, adopted in many fields. To adopt newly developed technology, nuclear industry requires rigorous V and V procedure and tests to assure reliable operation. Adoption of digital system requires verification and testing of the OS for licensing. Commercial real-time operating system (RTOS) is targeted to apply to various, unpredictable needs, which makes it difficult to verify. For this reason, simple, application-oriented realtime OS is developed for the nuclear application. In this work, we show how to verify the developed RTOS at each development lifecycle. Commercial formal tool is used in specification and verification of the system. Based on the developed model, software in C language is automatically generated. Tests are performed for two purposes; one is to identify consistency between the verified model ...

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This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the ...

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This project is to provide the regulatory direction of containment bypass during multiple steam generator tube failure issue for the Korean Next Generation Reactors, which is a part of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : the Multiple Steam Generator Tube Repture(MSGTR) event has never been occurred in the history of commercial nuclear reactor operation but single Steam Generator Tube Rupture(SGTR) event is reported to occur every two years. A probabilistic safety analysis study on MSGTR event, however, show its probability of occurrence is to be the same order as the design basis accidents such as LACA. In this regard, the ability of NPPs to cope with MSGTR event is required. Some requirements on initial and boundary conditions are suggested to be used in the analyses of NPPs during MSGTR events. The items that should be considered in establishing regulatory ...

International Nuclear Information System (INIS)

A Real-Time Operating System(RTOS) is an Operating System(OS) intended for real-time applications. Benchmarking is a point of reference by which something can be measured. The QNX is a Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available Personal Computer(PC) for applications such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS). The ELMSYS PC Hardware is being qualified by KTL(Korea Testing Lab.) for use as a Cabinet Operator Module(COM). The QNX RTOS is being dedicated by Korea Atomic Energy Research Institute (KAERI). This paper describes the outline and benchmarking test results on Context Switching, Message Passing, Synchronization and Deadline Violation of QNX RTOS under the ELMSYS PC platform

International Nuclear Information System (INIS)

A biodegradable polylactic acid (PLA)/poly(glycolide-co-lactide) copolymer (PLGA) membrane with polyglycolic acid (PGA) mesh was prepared to aid the effective regeneration of defective periodontal tissues. The microporous membrane used in this study consists of biodegradable polymers, and seems to have a structure to provide appropriate properties for periodontal tissue regeneration. Based on the albumin permeation test, it is known that the biodegradable membrane exhibits the suitable permeability of nutrients. The membrane maintained its physical integrity for 6-8 weeks, which could be sufficient to retain space in the periodontal pocket. Cell attachment and cytotoxicity tests were performed with respect to the evaluation of biocompatibility of the membrane. As a result, the membrane did not show any cytotoxicity. The safety and therapeutic efficacies of the biodegradable membranes were confirmed in animal ...

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The Dowex 50W-X8 and 50W-Xl2 resin samples are both strong acid cation materials in the hydrogen form. Each material has a water retention capacity characteristic of its respective marketed degree of cross-linking. Dowex 21K gives confirmatory responses to tests for a strong anion exchange resin in the nitrate form. All three resins have the manufacturer's specified ionic type and form, and the Dowex 50W resins have characteristic water retention capacities. These tests conclude that the ion exchange resins in use in FB-Line meet the approved safety document criteria for cross-linking, ionic form, and resin type.

International Nuclear Information System (INIS)

The Cycle 8A static tests conducted in the Fast Flux Test Facility (FFTF) during 1986 have resulted in the separation of various feedback reactivity components. These feedback components, described by closed-form equations depending only on the reactor temperature field, can be regarded as database for the validation and/or calibration of feedback mechanistic models. The SASSYS safety analysis code contains the most developed feedback reactivity models and was selected for the comparison study between database and mechanistic calculations for the FFTF. Although detailed feedback models for control rod repositioning and core radial expansion/bowing exist, only the simple models were available in SASSYS at the time of this study. The results are described in this paper.

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

The purposes of the pipe rupture studies at the Japan Atomic Energy Research Institute are to perform the model tests on the pipe whip of a pipe-restraints system, to get jet impingement force and blowdown thrust force, and to establish the computational method for the analysis of these phenomena. This paper presents the experimental and analytical results of the pipe whip tests carried out under the PWR LOCA conditions using the test pipe of 4-inch diameter and the U-shaped restraints. In the tests, the gap between the test pipe and the restraints was set nearly constant and the overhang length was 250 mm, 400 mm or 650 mm. The dynamic strains and residual deformations of the test pipe and restraints, and the restraint force were measured to clarify the effects of the overhang length on the pipe whip behaviors of the pipe-restraints system. It was confirmed ...

International Nuclear Information System (INIS)

Over the past few years, NPCIL has performed comprehensive Level-1 Probabilistic Safety Assessment (PSA) for a 220 MWe Pressurised Heavy Water Reactor (PHWR) at Kakrapar Atomic Power Station (KAPS) and for the first 540 MWe PHWR at Tarapur Atomic Power Project (TAPP- 3 and 4). The major objective of these PSAs was to present an integrated picture of the safety of the plant to identify and understand key plant vulnerabilities. As a result of the availability of these PSAs, there is a desire to use them to operate the plants in the most efficient manner practicable. In recent years, the operation of Indian Nuclear Power Plants has been characterized by improved availability/capacity factors and reduced forced outages. Frequency of planned outages is also being reduced. In order to achieve this, the PSAs are now being used as an engineering tool for optimization of Technical Specifications with regard to Allowed Outage Time (AOT) and Surveillance ...

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The DGs (Diesel Generators) in NPP (Nuclear Power Plant) has been used for the emergency electric power source to shot down the nuclear reactor safety in case of station blackout. The RCM (Reliability Centered Maintenance) has been applied to DGs for increasing the safety of NPP. The structured defects of DG were not remedied by the improvement of maintenance method. As the first stage of RCM, to find the structured defect, its failure= models were searched and analyzed through the ten year maintenance information. The structured defects such as the air compressor, the lubricating oil pressure, and the insufficient load were the root causes of main failures. The air reservoir reinstallation, the lubricating oil tube modification, the load bank installation, and the qualitative instrumentation were the solutions for the hardware oriented RCM of DGs. There remains the software oriented RCM such as the rejection of useless ...

Energy Technology Data Exchange (ETDEWEB)

The application of the LBB technology to the newly constructed PWRs has been approved in Korea for several high energy systems that can meet rigorous acceptance criteria. The LBB evaluation had been initiated for Yong-Gwang 3 and 4 units (YGN 3 and 4) which are in service, and subsequently performed for Ul-Chin 3 and 4 units (UCN 3 and 4) which are in construction. The LBB application in Korea is based on the USNRC requirements. The LBB evaluation for YGN 3 and 4 is to eliminate the dynamic effects associated with the postulated DEGB from design basis loads as well as to eliminate pipe whip restraints and jet impingement barriers. There were several issues on the application of LBB to the primary coolant loop and the pressurizer surge line. Of concern were material properties of carbon steel for the primary coolant loop, estimation of crack opening area at the pipe-to-nozzle interface considering the asymmetry, and the leakage crack size which barely meets the ...

International Nuclear Information System (INIS)

LOCA (Loss of coolant accident) resistant heat-shrinkable polymeric products are widely used for the connection of LV cables for class 1E systems inside the containment area of nuclear power plants. The paper/presentation describes the verification of a reformulated compound for these products, where certain components of the compound formulation had to be substituted. A qualification programme with this so-called reformulated compound was undertaken to proof the equivalency of the products to the products made out of the original compound. The basic elements of this requalification project were material qualification tests including accelerated aging tests according to the arrhenius method and type tests in accordance to IEEE 383, including flammability tests. The test results showed that the products made out of the reformulated compound were similar in fit, form and function to ...

Science.gov (United States)

Propionibacterium freudenreichii ET-3 culture, a cell-free product of whey fermentation using P. freudenreichii ET-3 (7025), has been shown to promote the growth of Bifidobacteria through the action of 1,4-dihydroxy-2-naphthoic acid (DHNA), and therefore, has potential use in the food and supplement industries. Although currently used as a food ingredient in Japan, the safety of this novel ingredient has not been previously evaluated through traditional toxicity testing. Therefore, here we report the results of standard toxicological testing performed on P. freudenreichii ET-3 culture. In a 4-week oral toxicity study, administration of 6000mg/kg body weight/day P. freudenreichii ET-3 culture was without compound-related adverse effects on clinical signs, body weights, food consumption, ophthalmology, hematology, clinical chemistry, urinalysis, organ weights, and gross and microscopic findings in male and female ...

British Library Electronic Table of Contents (United Kingdom)

A test bench was designed and assembled to carry out impact tests on samples and components. The system allows simple and rapid adjustment of the test parameters, such as the shot size and air pressure, with good repeatability of the results. Tests on steel and light alloys were carried out under both as-produced condition and on thin hard-coated samples. Significant reductions in dimple dimensions were seen after coating. FE models simulating the experiments overestimated the dimple depths, although the parameter trend was satisfactorily captured. The residual stresses from coating and shot peening determined numerically are believed to have been proven effective against fatigue.

Energy Technology Data Exchange (ETDEWEB)

This study is related to the use of thermoplastic liners to control corrosion of new and rehabilitated pipelines carrying aggressive hydrocarbon products. The aim is to investigate the technical and safety issues involved in introducing a thermoplastic liner into an existing carbon steel pipeline for aggressive hydrocarbon service at temperatures up to 150 C and internal pressures up to 34 MPa (5,000 psi). The principle objective is to provide design data for thermoplastic lined pipelines. The scope combines the use of liner insertion techniques with thermoplastic materials widely used in onshore pipelines and demonstrates the effectiveness of thermoplastic linings. The results indicate that thermoplastic liners can be used to control corrosion of pipelines in aggressive hydrocarbon environments.

Energy Technology Data Exchange (ETDEWEB)

The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events.

Energy Technology Data Exchange (ETDEWEB)

Nuclear generation operators are facing a challenging business environment at the beginning of the new millennium. Evolving changes in business context, competitive commercial pressures, and changes in technology have dictated recurring evaluation of operational practices and the adequacy of supporting tools, and the pursuit of opportunities for operational improvement. A key area of utility operations that has been impacted by these changes is the nuclear plant control centre. Changes to workspace layout, equipment provisions, staffing, and work organization are examples of some of the adjustments being introduced to improve operational and safety effectiveness. This paper discusses some of the key factors influencing these changes and identifies additional design principles for CANDU control centres that will enable new control centre designs and retrofits of existing control centres to remain relevant and responsive to utility needs. ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

In order to build the safety culture for nuclear power industry, it is important to evaluate the safety culture scientifically. Considering the traits of safety culture in the nuclear power industry, 24 safety culture assessment indexes are established from 4 aspects such as Safety consciousness, Safety attitude, Safety action and Safety actuality by using the SMART criteria. Safety culture star-class assessment criterion is presented and safety culture star-class assessment system is developed by using Visual Basic 6.0 and BP neural network. The system has a better generalization ability, and it can show exactly which phase the safety culture is in. Experimental results show that safety culture star-class assessment is practical and easy ...

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)

Geological clay formations are investigated for use as final underground deposit for heat producing nuclear waste. Special kinds of clay (e.g. bentonite) can also be used for the construction of geotechnical barriers. For the long time safety prognosis of the nuclear waste repositories the development of geochemical transport models is indispensable. The transport of aqueous solutions in clay is a complex process. The three-layer-minerals bentonite and illite swell by the adsorption of water, if the volume is restricted a high swelling pressure develops. The excellent barrier effect of natural clay formations and geotechnical clay barriers is based on the high swelling pressure and the high adsorption capacity for radionuclides and other pollutants. The two-layer-mineral kaolinite has no swelling capacity. In contrast to sandy layers a special geochemical transport potential exists in clay besides the well known matrix ...

Energy Technology Data Exchange (ETDEWEB)

The excellent high temperature strength and thermal conductivity of molybdenum-base alloys provide attractive features for components in advanced magnetic and inertial fusion devices. Refractory metal alloys react readily with oxygen and other gases. Oxidized molybdenum in turn is susceptible to losses from volatile molybdenum trioxide species, (MoO3)m, in air and the hydroxide, MoO2(OH)2, formed from water vapor. Transport of radioactivity by the volatilization, migration, and re-deposition of these volatile species during a potential accident involving a loss of vacuum or inert environment represents a safety issue. In this report we present experimental results on the oxidation, volatilization and re-deposition of molybdenum from TZM in flowing air between 400 and 800°C. These results are compared with calculations obtained from a vaporization mass transfer model using chemical thermodynamic data for vapor pressures of MoO3(g) over ...

International Nuclear Information System (INIS)

On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract ...

International Nuclear Information System (INIS)

The paper describes the experimental and analytical results of the pipe whip tests performed under the PWR LOCA conditions using 4, 6 and 8 inch test pipes. The tests were carried out at an initial pressure and a temperature of 15.7 MPa and 325 "0C. Two different types of tests were performed. One was the cantilever type pipe whip test using the test pipe of 3000 mm in length and U-shaped restraints. The other was the cross-over leg pipe whip test using a 1/6 model of piping in the PWR nuclear power plants. The cantilever type pipe whip tests were performed to investigate the influences of overhang length and pipe diameter on the pipe whip behavior. The movement of the test pipe is limited effectively by the restraints when the overhang length is short. The restraint force ...

Energy Technology Data Exchange (ETDEWEB)

Using a facility named SPHINX, which can accommodate a heat transfer test with CO{sub 2} at supercritical pressure, a series of tests was performed. The test geometries include tubes with the inner diameter of 4.4, 6.32 and 9 mm. a concentric annular passages with 8 x 10 mm, and an eccentric annular passages with 9.5 x 12.5 mm. Based on the test results, heat transfer correlations were developed and compared with the existing correlations. The heat transfer deterioration which may occur at certain conditions of heat and mass flux, were carefully studied and the published criteria were reviewed against our test results. Numerical calculation by using commercial CFD code, Fluent, were performed in order to provide the pre-test information for the heat transfer tests. Various turbulence models were evaluated and reliable ...

Energy Technology Data Exchange (ETDEWEB)

In this report, the descriptions of the DIVA test facility and those test matrixes are presented. The test sections simulate the downcomer annulus of DVI system with various scalling methodologies. The UPTF one has 1/7.5 scale and the APR-1400 test sections are reduced in 1/7 scale and 1/5 scale respectively. Also, the instrumentations, measuring methods are described. The major measuring parameters are ECC injection rate, air injection rate, bypass fraction, system pressure and so on. And finally, the operation procedure and experimental steps are introduced. From the DIVA tests, the multi-dimensional behaviors of two-phase flow in the downcomer during the reflood phase can be investigated using a transparent test facility. Also comparing the experimental data which are performed in various scale downcomer, it is expected that some criteria ...

Energy Technology Data Exchange (ETDEWEB)

By setting up a fully indicated, modern passenger car diesel engine, Mahle has created the basis for systematically testing current and future engine technologies. The experimental set-up has both an integrated high- and low-pressure EGR circuit and an open engine control unit, thus enabling efficient development of new systems under real-world conditions. The advantages of this comprehensive approach are illustrated below, using the example of different EGR technologies tested. (orig.)

International Nuclear Information System (INIS)

Plans, engineering parameters, and some test results for several FFTF core and primary sodium circuit instrument systems are presented. The systems discussed include temperature, flow, pressure, leak detectors, level sensors, fuel failure monitoring, sodium impurity analysis and cover gas monitors. Since many of these instruments are similar to those used in other fast reactors around the world, only a brief description is presented for these systems. Results of recent demonstration tests of the FFTF Under-Sodium Viewing and Ranging system are also presented. (U.K.).

Energy Technology Data Exchange (ETDEWEB)

The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces were obtained by Navier-Stokes momentum equation for a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained: (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant ...

Energy Technology Data Exchange (ETDEWEB)

Machine condition monitoring (MCM) has the capability to predict equipment maintenance needs which can reduce forced down-time and facilitate the avoidance of catastrophic failures and the consequential secondary damage. It can be used to allow confident deferral of routine maintenance, and improve equipment performance, availability, reliability and safety. In order to capitalize on the potential benefits of MCM directed toward power transformers two types of vibration tests were conducted on transformers owned and operated by Cargill Inc. and IES Utilities at the Bridgeport corn plant in Eddyville, Iowa. These tests involved collecting vibration signals from the transformer tank walls during transformer energization at no-load and during steady state operation at various loads. These vibration signals were then used to detect deterioration of internal mechanical integrity. Both types of tests ...

International Nuclear Information System (INIS)

Experimental and theoretical work have been carried out to find out the effects of heat transfer augmentation on two-phase flow instabilities in a single channel system. The effect of mass flow rate, heat input and inlet subcooling on the system behavior is studied using Freon-11 as the test fluid and six different heater surfaces are tested at various experimental conditions. Experimental evaluation is done using the steady-state pressure drop versus mass flow rate curves along with the curves of additional inlet pressure drop required to stabilize the system during the oscillations, and tables generated using the experimental data. Homogeneous equilibrium flow model and finite differences are used in theoretical analysis. An empirical relationship is obtained to calculate the pressure drop across the exit restriction. Characteristic equation for the system is found by linearizing ...

Energy Technology Data Exchange (ETDEWEB)

This semiannual report for the period October 1983-March 1984 summarizes activities in Sandia National Laboratories' continuing program of coal liquefaction research. The primary goals are to: explore novel catalytic concepts and materials for conversion of coal to liquid fuels; determine the effects of process variables on catalyst deactivation; determine the effects of coal structure and solvent properties on low temperature dissolution; study the kinetics and catalysis of hydrogen transfer reactions; develop an understanding of slurry gelling phenomena; and provide a technical assessment of coal liquefaction processes. During this period, work was performed on: the use of pyrene as a chemical probe of catalyst activity; analysis of catalysts from Wilsonville run 242 using ESCA; atmospheric pressure model compound activity testing of regenerated catalysts from Wilsonville run 242; base displacement experiments with a coal-indole ...

Energy Technology Data Exchange (ETDEWEB)

This paper summarizes results of reviews on the safety, critical software performed during the licensing process for the new Wolsung units. Each of these CANDU-type nuclear power plants has two micro-computerized shutdown systems. The SDS No. 1 program is graphically programmed in such a manner that its development process does not essentially differ from the design process of the conventional analog counterpart. This approach is understandable even to a reviewer in the regulatory agency without additional training in software engineering. The confidence in the reliability of this system is strengthened by the reverse verification and increased by extensive testing such as the reliability test. Concerning SDS No. 2, the development process is significantly influenced by the {open_quotes}software cost reduction project{close_quotes} of the U.S. naval research laboratory, and is, as a whole, clear and well structured except ...

Science.gov (United States)

The Heavy Water Reactor Facility (HWRF), as part of the defense-in-depth philosophy to mitigate the effect of design-basis and severe accidents, is equipped with a passive containment cooling system (PCCS). The function of the PCCS is to provide a safety-grade path to the ultimate heat sink for the removal of the reactor coolant system sensible heat and core decay heat. Ambient air enters an annular space between the steel containment shell and the surrounding concrete shield building through inlets in the shield building wall, is heated via natural convection, rises, and exits the building through a chimney located above the containment dome. A test program is in place to access parameters important to the effective operation of the PCCS. This paper focuses on the large-scale tests (LSTs). The objectives of these tests are as follows: (1) demonstrate natural circulation cooling with more prototypic ...