British Library Electronic Table of Contents (United Kingdom)

To estimate the success criteria of an operators action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the `ASME PRA standard' also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called ...

International Nuclear Information System (INIS)

The process of ensuring the quality of computer codes can be very time consuming and expensive. The Oak Ridge National Laboratory (ORNL) Monte Carlo codes all predate the existence of quality assurance (QA) standards and configuration control. The number of person-years and the amount of money spent on code development make it impossible to adhere strictly to all the current requirements. At ORNL, the Nuclear Engineering Applications Section of the Computing Applications Division is responsible for the development, maintenance, and application of the Monte Carlo codes MORSE and KENO. The KENO code is used for doing criticality analyses; the MORSE code, which has two official versions, CGA and SGC, is used for radiation transport analyses. Because KENO and MORSE were very thoroughly checked out over ...

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the 'ASME PRA standard' also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS ...

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)

Energy Technology Data Exchange (ETDEWEB)

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

International Nuclear Information System (INIS)

A lot of effort has been made to spread the application of RTOS in embedded system development in China. RTOS, #mu#C/OS and #mu#C/OS-II, with its opened source code, is well known in embedded systems. Learning RTOS, Using RTOS, becomes more and more popular. The Book of 'Micro C/OS-II The Real-Time Kennel Second Edition' was published in 2002. Here some new progress of the RTOS, especially the safety certifications, the reliability of software is introduced. (authors)

International Nuclear Information System (INIS)

With regard to the operation of nuclear power plants, Electricite de France formally recognized in 1973 that it was necessary for safety reasons and economically acceptable to adopt organizational principles of quality assurance that would be applicable both to its own activities and to those of its suppliers. Generally speaking, the form and spirit of the quality assurance programme chosen is based largely on the Code of Practice No. 50-C-QA. In particular, the programme focuses on the flexible character of quality assurance requirements and stresses that in the final analysis product quality depends above all on those to whom the project has been assigned, because it is they who are responsible for meeting the quality objectives set. Ten years of experience with the suppliers of Electricite de France has shown that these suppliers, after some initial difficulty, have been able to adapt the application ...

Energy Technology Data Exchange (ETDEWEB)

The primary scope of this study is to establish the analysis technology for the advanced reactor designed on the basis of the passive and inherent safety concepts. This study is extended to the application of these technology to the safety analysis of the passive reactor. The study was performed for the small and medium sized reactor and the large sized reactor by focusing on the development of the analysis technology for the passive components. Among the identified concepts the once-through steam generator, the natural circulation of the integral reactor, heat pipe for containment cooling, and hydraulic valve were selected as the high priority items to be developed and the related studies are being performed for these items. For the large sized passive reactor, the study plans to extend the applicability of the best estimate computer code RELAP5/MOD3 which is widely used for the ...

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 ...

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 ...

International Nuclear Information System (INIS)

Nuclear facilities of the Department of Energy (DOE) located at the Savannah River Site must comply with DOE orders as implemented at DOE-SR. The DOE orders cover safety criteria, design criteria, environmental protection, occupational health and safety. The program applies to DOE and contractors. In this section, the Nuclear Regulatory Commission (NRC) criteria and industry codes and standards are addressed as well as DOE orders. Specific DOE orders which add additional criteria have also been noted. A program for assessing and implementing contractor applicable DOE orders has been established. This program ensures that compliance is achieved through developing and implementing policies, programs, and procedures. The primary emphasis is placed on safe, efficient reactor restart and operation. DOE has classified orders applicable to restart as Level I, Category I while those ...

Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: The Nuclear Power Plants (NPPs) are always designed for the highest level of safety against postulated accidents which may be initiated due to internal or external causes. One of the external/internal causes, which may lead to accident in the reactor and its associated systems, is fire in certain vital areas of the plant. Conventionally, the fire containment approach and/or the fire confinement approach is used in designing the fire protection systems of NPPs. Indian NPPs (PHWRs) follow the combined approach to ensure plant safety and all newly designed plants are required to comply with the provisions of Atomic Energy Regulatory Board (AERB) fire safety Guide. In respect of older plants, the reassessment of adequacy of fire safety provisions in the light of current advances has becomes essential so as to decide upon the steps for retrofitting. Keeping this in mind the ...

International Nuclear Information System (INIS)

National Nuclear Safety Department is authorized by Infra (Iranian Nuclear Regulatory Authority) for issuing rules and regulation and conducting the licensing and supervisory process for nuclear facilities. The main responsibilities of the NNSD are conducted via five main division are, nuclear codes and standards, Nuclear Safety assessment, Authorization, Inspection and Enforcement and Technical Support. In this paper, the functions and responsibilities of TSD in enhancing nuclear safety are described. Examples of main tasks to support the Nuclear Safety assessment division regarding to technical calculations and research of safety issues in licensing documents are provided. (author)

International Nuclear Information System (INIS)

This paper will review code and standard and the safety related features of major components of Monju: Components of the Reactor Coolant Boundary; Components of the Reactor Shurdown Systems; Components of the Decay Heat Removal Systems; Components of the Engineered Safety Features; Other Safety Related Components. Their relationship to the system or plant function is emphasized, in reviewing these components.

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 ...

International Nuclear Information System (INIS)

A nuclear power plant engineering (NPP) simulator may include many system functions such as thermal-hydraulic calculation, 3D reactor neutron kinetics model, control and protection system, display and operational human-machine interface, intelligent alarm system, etc. If all those functions are simulated using a single code, the size of the program will cause structural or managerial problems, and the hardware requirement could be tremendous. Thus, any local function errors or changes may affect the whole code, which will make development and maintenance extremely costly. So the nondistributed code is not considered to be flexible and feasible. Currently, distributed and parallel technologies have been applied to develop NPPs engineering simulator for safety analysis, verification of advanced main control room display, operation and intelligent alarm design. The distributed simulation divides a huge ...

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 and the generated ...

International Nuclear Information System (INIS)

Full text of publication follows: The TUV NORD e.V. is an independent Technical Support Organisation (TSO) performing safety assessments in almost every field of technology. In nuclear safety the TUV can look back on more than 40 years of experience. In the last years in Germany PWR safety analyses were focussed on boron dilution events with the potential of reactivity transients. The possibility of coolant with a low boron concentration collected in localized areas of the reactor coolant system (RCS) can be caused by injection of coolant with less boron content from interfacing systems (external dilution) as well as separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution). Inherent dilution can e.g. occur after reflux-condenser heat transfer after a small break loss of coolant accident (SBLOCA) with a limited operability of the emergency core cooling (ECC) systems. The TUV ...

Energy Technology Data Exchange (ETDEWEB)

An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. ...

International Nuclear Information System (INIS)

The applicability of the thermal response of an electrically heated simulated rod mostly used in loss-of-coolant-accident (LOCA) experiments to that of a nuclear fuel rod is a concern for the safety evaluation of a reactor. The present analysis describes the characteristics of the thermal response for both electrically heated and nuclear fuel rods during typical reflood conditions for a PWR-LOCA. A model describing the radial temperature field in the rod is developed based on the scheme in HETRAP code by Malang and incorporated into a reflood analysis code, REFLA for that purpose. The calculations applied to the existing reflood tests gave good agreement with experiments, showing the validity of the present model. The analysis has shown that the nuclear fuel rod tends to give a lower clad temperature and a sooner quench time than the electrically heated rod in a typical reflood condition, due to the ...

Energy Technology Data Exchange (ETDEWEB)

Piping in nuclear power plants is vital to the proper operation and safety of these facilities. To assure safety in the unlikely event of a pipe break, it is necessary to evaluate the consequences from the resulting whipping pipe on neighboring components and structures. Numerical simulations allow for rapid evaluation of these consequences. Before simulations can be accepted, however, the methodology and computer codes must be validated against experimental results. This paper uses a probabilistic approach to validate pipe whip simulations against limited experimental results. Probabilistic analysis software was developed and coupled to existing deterministic finite element software. An example of a whipping pipe impacting against a reinforced concrete slab was simulated. The described probabilistic approach was used to validate the numerical simulations. The conclusions obtained were that the numerical simulations of ...

Energy Technology Data Exchange (ETDEWEB)

Turret moored Floating Production Storage and Offloading (FPSO) systems are considered suitable for a wide range of environmental conditions but experience of designing moorings for these units in harsh environments is limited. Furthermore, each floating production system has its own requirements as to the way the riser system is operated because of its dependence on both reservoir and site specific conditions. The current mooring codes do not have a consistent set of requirements; for example, the intact mooring system safety factors range from 1.5 to 2.3. Such vastly differing requirements result in wide variability in the intrinsic safety, reliability and cost. This paper describes a reliability analysis carried out for the first North Atlantic FPSO (Foinaven), to address the acceptance criteria for moorings by returning to the fundamentals. Although the analysis deals with FPSO`s, the approach is ...

International Nuclear Information System (INIS)

In this paper, we put forward a systematic method to analyze, control and evaluate the safety issues of medical robotics. We created a safety model that consists of three axes to analyze safety factors. Software and hardware are the two material axes. The third axis is the policy that controls all phases of design, production, testing and application of the robot system. The policy was defined as hazard identification and safety insurance control (HISIC) that includes seven principles: definitions and requirements, hazard identification, safety insurance control, safety critical limits, monitoring and control, verification and validation, system log and documentation. HISIC was implemented in the development of a robot for urological applications that was known as URObot. The URObot is a universal robot with different ...

Energy Technology Data Exchange (ETDEWEB)

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW ...

Energy Technology Data Exchange (ETDEWEB)

Full text of publication follows: During calculations of SAR accidents and transients it is necessary to perform steam generator simulation. Best accuracy is 3D transient calculations presented in report. Main outcomes of work was next: 1. There was shown by analysis the applicability of code TRAC (Los-Alamos laboratory) for thermal - hydraulic calculations of horizontal steam generator PGV-1000M. Special nodalization scheme was developed for it purposes. 2. Validation and selection of thermal-hydraulic correlations for improvement of using the code at calculation PGV-1000M were performed. As result Labuntsov formula is recommended for horizontal SG. 3. Calculations of nominal mode operation of PGV-1000M for cross-verification with code STEG (Electrogorsk Research and Engineering Center EREC) during its verification were performed. Solution by TRAC was obtained for transient problem after stabilization ...

International Nuclear Information System (INIS)

In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation ...

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

Energy Technology Data Exchange (ETDEWEB)

This report, the seventh in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1991. The goal of this office is to ensure that safety-related research provides the technical bases for rulemaking and for related decisions in support of NRC licensing and inspection activities. This research is necessary to make certain that the regulations that are imposed on licensees provide an adequate margin of safety so as to protect the health and safety of the public. This report describes both the direct contributions to scientific and technical knowledge with regard to nuclear safety and their regulatory applications.

Energy Technology Data Exchange (ETDEWEB)

This report, the sixth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1990. The goal of this office is to ensure that safety-related research provides the technical bases for rulemaking and for related decisions in support of NRC licensing and inspection activities. This research is necessary to make certain that the regulations that are imposed on licensees provide an adequate margin of safety so as to protect the health and safety of the public. This report describes both the direct contributions to scientific and technical knowledge with regard to nuclear safety and their regulatory applications.

Energy Technology Data Exchange (ETDEWEB)

Implementation of Title 10 Code of Federal Regulations Part 830, Subpart B Nuclear Safety Management (1) was accomplished at the Waste Isolation Pilot Plant (WIPP) in a timely and efficient manner. The primary reason the transition went smoothly was that the existing safety analysis was relatively new, initially developed in 1995, and written in accordance with the safe harbor document DOE-STD-3009 (2). The WIPP Safety Analysis Report (SAR) (3) is kept up-to-date with the unreviewed safety question (USQ) process and thorough oversight and input provided by DOE-Carlsbad Field Office (CBFO) documented in the annual safety evaluation report (SER) process.

Energy Technology Data Exchange (ETDEWEB)

The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.).

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of the verification test ...

International Nuclear Information System (INIS)

A combination of computer codes such as LATREP, HWRAXAV and CITATION is utilized in an attempt to analyze the nuclear design characteristics of the CAXDU-PHWR of the Wolsung Unit 1. The major nuclear properties to be computed are the lattice properties of CANDU fuel channel and the core channel power distribution. The computed results are compared with the preliminary safety reports documentation for the Wolsung reactor. The observed discrepancies between our computations and the preliminary safety reports values are discussed in terms of incomplete information on the description of the core configuration in the preliminary safety reports and the different calculation methods. (author).

Energy Technology Data Exchange (ETDEWEB)

In the thermal hydraulics codes developed for fire safety analysis and for containment thermal hydraulic analysis, junctions in the multi-compartment geometries is often modeled as uni-directional junctions. However, ceiling junctions are known to depict unstable/oscillatory bi-directional flow behavior. Detailed investigations have been carried out to understand the unstable flow behaviour of a junction by analyzing an earlier reported experiment and its subsequent two dimensional numerical RANS based study of fire in an enclosure. The authors attempt more realistic and desired three dimensional and inherently transient large eddy simulations using a computer code Fire Dynamics Simulator (FDS). The paper presents the details of the analysis, the results obtained and further studies required to be conducted so that the findings can be applied to the fire/containment thermal hydraulics analysis codes ...

Science.gov (United States)

Hamed, A. and A. Mohamed, "Assessment of Shock Induced Flow Separation and ...... AGARD Symposium on Combined Cycle Propulsion for Hypersonic Application, ...

Energy Technology Data Exchange (ETDEWEB)

A three-dimensional (3D) thermal-hydraulic model and a numerical procedure for the simulation and analysis of a steady-state, as well as transient operation of nuclear power plant components are presented. A two-fluid approach is applied to modelling of two-phase flow. Thermal-hydraulics of a horizontal steam generator in the WWER 1000 nuclear power plant has been simulated at the full load, steady-state operation. A comparison of the numerical results with data measured at the NPP Novovoronjezh shows good agreement. 3D numerical results can be used in plant design or retrofitting, in nuclear power plant operation and safety analysis and as improvement of existing one-dimensional thermal-hydraulics models of the horizontal steam generator which are assessed by system codes used for the nuclear power plant safety analyses. (author)

International Nuclear Information System (INIS)

From the beginning, one of the most useful applications of Probabilistic Safety Assessment (PSA) is its use in evaluating the risk importance of changes to plant design, operations, or other plant conditions. Risk importance measures the impact of a change on the risk. Risk is defined as a combination of the likelihood of failure and consequence of the failure. The consequence can be safety system unavailability, core melt frequency, early release, or various other consequence measures. The goal in this PSA application is to evaluate the risk importance of an ISI process, as applied to plant piping systems. Two approaches can be taken in this evaluation: Current PSA Approach or the Blended Approach. Both are discussed here.

Energy Technology Data Exchange (ETDEWEB)

Various physics applications involve the computation of the standard hypergeometric function {sub 2}F{sub 1} and its derivatives. Because it is not an intrinsic in the common programming languages, automatic differentiation tools will either differentiate through the code that computes {sub 2}F{sub 1} if that code is available or require the user to provide a hand-written derivative code. We present options for the derivative computation in the context of an ionization problem and compare the approach implemented in the Diamant library to standard methods.

Energy Technology Data Exchange (ETDEWEB)

This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P{sub N}) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results ...

CERN Document Server

This paper presents prefix codes which minimize various criteria constructed as a convex combination of maximum codeword length and average codeword length or maximum redundancy and average redundancy, including a convex combination of the average of an exponential function of the codeword length and the average redundancy. This framework encompasses as a special case several criteria previously investigated in the literature, while relations to universal coding is discussed. The coding algorithm derived is parametric resulting in re-adjusting the initial source probabilities via a weighted probability vector according to a merging rule. The level of desirable merging has implication in applications where the maximum codeword length is bounded.

International Nuclear Information System (INIS)

The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref.

International Nuclear Information System (INIS)

The aim of the safety calculation is to quantify through numerical modelling the radiological impact (molar flow, human dose) of a potential repository for radioactive waste on the Meuse/Haute Marne site at Bure. A selection process is underway for safety scenarios and their phenomenological and numerical conceptual models upstream from the safety calculation. This involves defining and quantifying the geometric and dimensional representations of the repository for each scenario plus the physical, mathematical and numerical models that reflect its behaviour and the uncertainties associated with all the parameters required to quantify the impact. A summary will be given of these various aspects. The numerical simulations are then performed on the Alliances platform which integrates the various computer codes required for the physical representation of the system. (authors)

International Nuclear Information System (INIS)

The Decree specifies basic quality assurance requirements applicable to machines, their parts and materials, civil engineering structures, means for automated control of technological processes including hardware and software, and electricity supply systems related to nuclear safety of nuclear facilities, and stipulates binding procedures for the implementation of technical and organizational provisions associated with the quality of selected equipment to ensure nuclear safety of nuclear facilities. Safety classes are defined for selected equipment. Requirements laid on safety assurance documentation are specified, and requirements placed on safety assurance programmes, their preparation, finalization and approval are defined. Quality assurance requirements are also specified with respect to the designing, manufacture, construction, operation, repair, ...

Energy Technology Data Exchange (ETDEWEB)

With the recent advance of the utilization of nuclear energy in a large scale, social interest is being focussed in the potential risk which the nuclear technology will accompany. Especially after the accidents in Chernobyl and other nuclear facilities, serious anxiety to the utilization of nuclear energy is prevailing among the general public. In order to meet the anxiety and distrust of the population in the use of the nuclear power, the health effect or risk which radioactive materials released into the environment will bring about should be comprehensively and properly evaluated, and then should be widely reported to the population. The development of HESANS code system (Health and Environmental Safety Assessment Network System) was planned to set up such a comprehensive computer code that covers a whole pathway of radioactive material from its release to estimates of derived health effects in the population, including ...

Science.gov (United States)

FIRESET is a PC-based computer code which calculates current as a function of time for an RLC circuit containing up to fifteen series conductors which undergo rapid heating and subsequent explosion as a consequence of an electric current which passes through them. In its original form, the code was developed to model electrical waveforms measured when a large, typically 25.4 x 25.4 x 0.051-mm, aluminum foil was exploded using a capacitor bank with tens of kilojoules of stored energy. The code proved to be useful for this purpose, and it was recognized that it could also be used for modeling the electrical response of detonator bridgewires. In view of the increasing use of slapper detonators for DOD applications, we wish to make the latest version of the code, available to DOD laboratories and contractors for use in designing firing systems which employ slapper or exploding ...

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)

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.

International Nuclear Information System (INIS)

Wolsong 2,3 and 4 fuel channel analysis during a large break loss of coolant accident with loss of ECCS injection (LOCA/LOECC) is performed to obtain the heat load to moderator. Because the single channel analysis requires the establishment of the safety codes and their input decks, the present study follows the same safety analysis methodology found in FSAR of Wolsong 2,3 and 4. From this work we obtain the safety tools such as CATHENA MOD3.5b/Rev.1 and CHAN-II/A MOD2 codes, and their code modeling in a form of code input deck. The analysis consists of two parts: front-end (blowdown period) and back-end. For the front-end analysis the fuel and pressure tube (PT) temperatures, and PT circumferential strains at the end of front-end as well as fuel channel depressurization are calculated using CATHENA code and used as ...

Energy Technology Data Exchange (ETDEWEB)

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW ...

CERN Document Server

This paper presents an application of the state of the art and new trends for risk management of safety-related control and monitoring systems, currently applied in the industry. These techniques not only enable to manage safety and reliability issues but they also help in the control of quality and economic factors affected by the availability and maintenance of the system. The method includes an unambiguous definition of the system in terms of functions and a systematic analysis of hazardous situations, undesired events and possible malfunctions. It also includes the identification and quantification of the risk associated to the system. The required risk reduction is specified in terms of safety integrity levels. The safety integrity level results in requirements, preventive measures, possible improvements and recommendations to assure the satisfactory management of the risk.

CERN Document Server

In the present study, applications of the SPH method to industrial related issues are considered by starting from an existing open source 2D SPH code, namely the SPHYSICS code, which offers an effective ground for numerical developments, which are performed in order to bring an answer to industrial problems, such as simulations of solid/fluid coupling in a free surface flow context. The purpose of the present paper is therefore to expose the numerical developments which yield an enhanced version (referred to as "SPHYSIC2") of the initial code. Firstly, the different features added to obtain the operational code needed for engineering applications are described, and so are the problems raised on this way, offering a kind of review of SPH methods for engineers. Secondly, the validation of the proposed code is partially presented with two well ...

Energy Technology Data Exchange (ETDEWEB)

A safety analysis code to design a pressurized water reactor and to obtain the licences including entire proprietary rights is under development in domestic research and development project. The purpose and scope of this report is to develop the flow regimes related models for inter-phase friction, wall frictions, wall heat transfer, and inter-phase heat and mass transfer in two-phase three-field equations. In order to choose choose the flow regime criteria, we have investigated various exiting best-estimate T/H codes in this chapter 2. They are the RELAP5-3D, TRAC-M, CATHARE, MARS codes. Around 500 references used in these codes have been collected and reviewed. Also we have investigated eleven papers in detail. In chapter 3, based on the selected flow regimes, the flow regime maps for a gas-liquid flow in horizontal and vertical tubes have decided including the mechanisms of flow ...

Energy Technology Data Exchange (ETDEWEB)

As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN (Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations for the thermohydraulic system analysis code RELAP5/MOD2/36.05. This document describes the assessment of this computer program versus a natural circulation experiment as conducted at the Borssele Nuclear Power Plant. The results of this comparison show that the code RELAP5/MOD2 predicts well the natural circulation behaviour of Nuclear Power Plant Borssele.

International Nuclear Information System (INIS)

Full text: Research reactors have been playing a multi dimensional role in areas of nuclear fuel cycle programme, radio-isotope productions, neutron beam research etc. To ensure an efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are required on routine basis. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation requires a prior estimation of the reactivity load due to the sample, heating rate and the activity developed in it during irradiation. For the safety of the personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr. Therefore, a proper ...

Wastenet

... My road safety has come a long way since a bus driver yelled 'Go and read the Highway Code!' at me in my first wobbly week. I've got over my bike theft paranoia and happily leave my trusty steed shackled to lampposts around London. And I've figured out the optimum skirt-length for pedaling - neither flasher short nor so long it gets stuck in the oily spokes. But ...

Science.gov (United States)

This thesis study covers both global performance and local phenomena analyses focusing on natural circulation reactor design safety. Four important topics are included: the global SBWR design safety assessment, important local phenomena investigation, steady and transient natural circulation process study, and two-phase instability analysis. The conceptual design of the SBWR-200 is introduced in this thesis and the global performance of a natural circulation reactor is then assessed using PUMA integral test data and RELAP5 simulations. A safety assessment methodology is developed to evaluate the PUMA integral test data extrapolation and code scalability. The RELAP5 code simulation capability in low-pressure low-flow conditions is also validated. The study shows that the code is capable of predicting the global accident scenario in natural circulation reactors ...

UK PubMed Central (United Kingdom)

We develop a statistical framework to study the relationship between chromatin features and gene expression. This can be used to predict gene expression of protein coding genes, as well as microRNAs....Full Text Available

International Nuclear Information System (INIS)

This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

International Nuclear Information System (INIS)

The days of high-temperature reactors in the Federal Republic of Germany are numbered. The AVR has been decommissioned, and an application has been filed for licensing the decommissioning of the THTR. Nevertheless, Prof. Dr. Rudolf Schulten who is the director of Juelich Nuclear Research Center's Institute for Reactor Development, and also full professor of Aachen Technical University in the field of reactor safety, predicts a good future for the HTR reactor line on a worldwide level, due to the inherent safety of this reactor type. (orig.).

International Nuclear Information System (INIS)

The method of syndrome coding for data compression read out from multiwire proportional chambers that has been previously proposed is generalized in case of its application to registration of the coordinates of events detected. The questions of execution of arithmetic and algebraic operations on the Galois field elements and their hardware implementation are considered. The method of computation is presented of a specialized processor for parallel computing the coordinates of three sparks. The estimate of its speed is equal to 185 ns. Data compression, data selection and coordinate calculations are performed without use of memory elements and timing pulses.

Energy Technology Data Exchange (ETDEWEB)

Inductive electromagnetic launchers, or coilguns, use discrete solenoidal coils to accelerate a coaxial conductive armature. To date, Sandia has been using an internally developed code, SLINGSHOT, as a point-mass lumped circuit element simulation tool for modeling coilgun behavior for design and verification purposes. This code has shortcomings in terms of accurately modeling gun performance under stressful electromagnetic propulsion environments. To correct for these limitations, it was decided to attempt to closely couple two Sandia simulation codes, Xyce and ALEGRA, to develop a more rigorous simulation capability for demanding launch applications. This report summarizes the modifications made to each respective code and the path forward to completing interfacing between them.

Energy Technology Data Exchange (ETDEWEB)

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of ...

Energy Technology Data Exchange (ETDEWEB)

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

The automotive industry desires to optimize safety and lightweight construction for the production of cars and trucks. The resulting task for steel suppliers is to make steel grades available that can fulfil these requirements. Different projects have shown optimistic outlooks for optimization of safety and reduction of the weight of automobile bodies by using different high strength and ultra high strength steels. In the last group the dual phase steels play the lead. In recent years different steel grades have been developed and optimized for these applications. On one hand developing the mechanical properties as well as joining and forming of dual phase steels is one of the first steps. On the other hand different coated surfaces for various applications of dual phase steels is another step. Starting with characteristic production conditions for cold rolled hot dip galvanised dual phase steels, the ...

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 ...

Science.gov (United States)

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

Energy Technology Data Exchange (ETDEWEB)

The main purpose of this report is to document the benchmarking results and verification of the RESRAD-OFFSITE code as part of the quality assurance requirements of the RESRAD development program. This documentation will enable the U.S. Department of Energy (DOE) and its contractors, and the U.S. Nuclear Regulatory Commission (NRC) and its licensees and other stakeholders to use the quality-assured version of the code to perform dose analysis in a risk-informed and technically defensible manner to demonstrate compliance with the NRC's License Termination Rule, Title 10, Part 20, Subpart E, of the Code of Federal Regulations (10 CFR Part 20, Subpart E); DOE's 10 CFR Part 834, Order 5400.5, ''Radiation Protection of the Public and the Environment''; and other Federal and State regulatory requirements as appropriate. The other purpose of this report is to document the ...

Energy Technology Data Exchange (ETDEWEB)

This paper identifies the event sequences that require thermal-hydraulic analyses for the success criteria of probabilistic safety analysis (PSA). The selection of the sequences is performed based on the review of the NEI Peer Review Process Guidance and ASME PRA Standard. Success criteria are the important element of PSA quality. Success criteria decide the success or failure of the key function in the PSA event tree. It is defined as a minimum set of components/trains of system required to mitigate an accident. Thermal-hydraulic codes are generally used to derive time-related criteria in the PSA, such as operator action time used in human reliability analysis (HRA), event timing, and time to recover the component or the power. This paper suggests the use of the MARS code for the T-H analysis to obtain the success criteria and sequence timing, and operator action time. In the Kori Units 3 and 4 PSA report, the T-H analyses ...

International Nuclear Information System (INIS)

The document represents a specific type of discussion of existing methodologies for the creation and application of probabilistic safety assessment (PSA) in light of the EUR document summarizing requirements placed by Western European NPP operators on the future design of nuclear power plants. A partial goal of this discussion consists in mapping, from the PSA point of view, those selected design, operational and/or safety factors of future NPPs that may be entirely new or, at least, newly addressed. Therefore, the terms of reference for this stage were formulated as follows: Assess current level of knowledge and procedures in the analysis of factors and phenomena with a dominant influence upon operational safety of new generation reactors, especially in the following areas: (1) Phenomenology of failure types and mechanisms and reliability of conventional passive safety system ...

British Library Electronic Table of Contents (United Kingdom)

This study deals with the neutronics analysis of the current core configuration of a 3MW TRIGA Mark-II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Safety Analysis Report (SAR) values. The comprehensive neutronics code system SRAC was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Cross-section data library generated from JENDL-3.2 were used. The validation of the model against benchmark experimental results is presented. The SRA...

International Nuclear Information System (INIS)

The present Manual provides guidance and illustrative examples for applying a method by which graded quality assurance requirements may be determined and adapted to the items and services of a nuclear power plant in conformance with the requirements of the IAEA Nuclear Safety Standards (NUSS) Code and Safety Guides on quality assurance. The Manual replaces the previous publication IAEA-TECDOC-303 on the same subject. Various methods of grading quality assurance are available in a number of Member States. During the development of the present Manual it was not considered practical to attempt to resolve the differences between those methods and it was preferred to identify and benefit from the good practices available in all the methods. The method presented in this Manual deals with the aspects of management, documentation, control, verification and administration which affect quality. 1 fig., 4 tabs.

Science.gov (United States)

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in ...

International Nuclear Information System (INIS)

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. ...

Energy Technology Data Exchange (ETDEWEB)

Recent modifications to the EQ3/6 geochemical modeling software package provide for the use of Pitzer's equations to calculate the activity coefficients of aqueous species and the activity of water. These changes extend the range of solute concentrations over which the codes can be used to dependably calculate equilibria in geochemical systems, and permit the inclusion of ion pairs, complexes, and undissociated acids and bases as explicit component species in the Pitzer model. Comparisons of calculations made by the EQ3NR and EQ6 compuer codes with experimental data confirm that the modifications not only allow the codes to accurately evaluate activity coefficients in concentrated solutions, but also permit prediction of solubility limits of evaporite minerals in brines at 25/sup 0/C and elevated temperatures. Calculations for a few salts can be made at temperatures up to approx. 300/sup 0/C, but the temperature ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

Methods are discussed for optimization of organizational model of longwall mining. Operations at a working face, their duration, number of miners, miner qualifications are analyzed. The so-called method for allocation of heteregeneous resources is used. Using the method optimum organizational models of working faces are constructed. The models consider working conditions, coal output, personnel, miner qualifications, duration of various operations and of idletime, and safety regulations. A computer code developed by the author for optimizing organizational models of longwall mining is described. 9 refs.

International Nuclear Information System (INIS)

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. ...

International Nuclear Information System (INIS)

To investigate the 2D/1D fusion core analysis method, a 1D neutron transport problem solver, PEACH-ID, is developed. It is a code of method of characteristics (MOC), both the usual fiat-source step characteristics (SC) scheme and linear source (LS) approximation scheme are adopted for tracking calculation along the neutron flying trajectory. Exponential function interpolation table and fission source extrapolation are adopted as two major methods to accelerate the computational process. Numerical results demonstrate that PEACH-1D is accurate and efficient, and the proposed LS scheme is able to handle quite larger mesh division and deserves much more application in the MOC codes. (authors)

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

The textbook is primarily intended for radiologic staff and radiologic safety officers and gives information on the current regulatory provisions of the German X-ray Ordinance, applications of X-rays, quality assurance, organisational aspects of film processing and quality requirements of X-rays. An annex lists the guidelines of the Bundesaerztekammer (German National Chamber of Physicians) relating to quality assurance aspects, and further useful information on commercially available film-screen systems, the various associations of physicians in Germany, and requirements and performance of radiation surveys. (vhe).

Energy Technology Data Exchange (ETDEWEB)

The days of high-temperature reactors in the Federal Republic of Germany are numbered. The AVR has been decommissioned, and an application has been filed for licensing the decommissioning of the THTR. Nevertheless, Prof. Dr. Rudolf Schulten who is the director of Juelich Nuclear Research Center's Institute for Reactor Development, and also full professor of Aachen Technical University in the field of reactor safety, predicts a good future for the HTR reactor line on a worldwide level, due to the inherent safety of this reactor type. (orig.).

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 ...

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

With the aid of a DOE grant (No. DE-FC26-01NT41050), Stolar Research Corporation (Stolar) developed the Horizon Sensor (HS) to distinguish between the different layers of a coal seam. Mounted on mining machine cutter drums, HS units can detect or sense the horizon between the coal seam and the roof and floor rock, providing the opportunity to accurately mine the section of the seam most desired. HS also enables accurate cutting of minimum height if that is the operator's objective. Often when cutting is done out-of-seam, the head-positioning function facilitates a fixed mining height to minimize dilution. With this technology, miners can still be at a remote location, yet cut only the clean coal, resulting in a much more efficient overall process. The objectives of this project were to demonstrate the feasibility of horizon sensing on mining machines and demonstrate that Horizon Sensing can allow coal to be cut cleaner and more efficiently. Stolar's primary goal was ...

International Nuclear Information System (INIS)

There are many differences between industrial CT and industrial radiography, such as imaging principle, inspection time, radiation dose and the requirements for operators etc. The national standards for radiation protection of industrial detection are not applicable to the requirements of protection and safety for #gamma#-ray industrial CT to some extent now. In order to standardize the production and use for #gamma#-ray industrial CT, protect the safety of operators and the public, and to promote the popularization and application of #gamma#-ray industrial CT, it is significant to establish the national standards for radiation protection of #gamma#-ray industrial CT as soon as possible. The purpose of this paper is to introduce the contents of this standard, and specify some important terms. Then there is a brief discussion on the existing problems during establishing such standards. At last, the paper ...

Energy Technology Data Exchange (ETDEWEB)

Coded imaging techniques using thick, micro-Fresnel zone plates as coded apertures have been used to image x-ray emissions (2-20 keV) and 3.5 MeV Alpha particle emissions from laser driven micro-implosions. Image resolution in these experiments was 3-8 ..mu..m. Extension of this coded imaging capability to higher energy x-rays (approx. 100 keV) and more penetrating charged particles (e.g. approx. 15 MeV protons) requires the fabrication of very thick (50-200 ..mu..m), high aspect ratio (10:1), gold Fresnel zone plates with narrow linewidths (5-25 ..mu..m) for use as coded aperatures. A reactive ion etch technique in oxygen has been used to produce thick zone plate patterns in polymer films. The polymer patterns serve as electroplating molds for the subsequent fabrication of the free-standing gold zone plate structures.

International Nuclear Information System (INIS)

The epithermal cross section shielding methods used in the lattice physics code EPRI-CELL (E-C) have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology. These include: treatment of the external moderator source with intermediate resonance (IR) theory, development of a new Dancoff factor expression to account for clad interactions, development of a new method for treating resonance interference, and application of a generalized least squares method to compute best-estimate values for the Bell factor and group-dependent IR parameters. The modified E-C code with its new ENDF/B-V cross section library is tested for several numerical benchmark problems. Integral parameters computed by EC are compared with those obtained with point-cross section Monte Carlo ...

Energy Technology Data Exchange (ETDEWEB)

Accurate calculations of atomic displacement damage in materials exposed to neutrons require detailed spectra for primary recoil nuclei. Such data are not available from direct experimental measurements. Moreover, they cannot always be computed accurately starting from evaluated nuclear data libraries such as ENDF/B-V that were developed primarily for neutron transport applications, because these libraries lack detailed energy-and-angle distributions for outgoing charged particles. Fortunately, a new generation of nuclear model codes is now available that can be used to fill in the missing spectra. One example is the preequilibrium statistical-model code GNASH. For heating and damage applications, a supplementary code called RECOIL has been developed. RECOIL uses detailed reaction data from GNASH, together with angular distributions based on Kalbach-Mann systematics to compute the ...

Energy Technology Data Exchange (ETDEWEB)

A couple of domestic institutions have been investigating the application of vitrification technology to treat low- and intermediate-level radioactive wastes in Korea. In the case that such investigations prove to be successful, it is expected that commercial vitrification plants will be constructed. The safety insuring on vitrification plants could not be compatible with criterion on radioactive waste management because the facilities are at high temperature and contain a variety of accommodations for the exhaust gases and residual products. Therefore, it is necessary to suggest a new strategy or modifications of criterion of radioactive waste management on considerations related with the vitrification technology. In order to ensure the safety of vitrification plants, a technical guideline or standard for design and operation of vitrification plants must be established too. A study on the safety ...

International Nuclear Information System (INIS)

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

International Nuclear Information System (INIS)

Treatment of food with low doses of gamma radiation has been recognized to have two main advantages. These consist of: (1) improvement of food safety by elimination of pathogens and (2) reduction of microbial spoilage and extension of shelf life of perishable items by reducing the number of viable spoilage organisms. Studies during the last few decades have conclusively proved the beneficial effects of radiation with respect to fishery products. The three potential areas of application to fish products include: (i) radurization for shelf life extension (ii) radicidation to eliminate food borne pathogens in the products and (iii) radiation treatment to dried products to control insects.

International Nuclear Information System (INIS)

125I-seeds are extensively used in ocular and interstitial brachytherapy for the treatment of various malignant lesions. Quality assurance and classification performance testing of indigenously produced 125I-seeds were carried out for ensuring their safety in different brachytherapy applications. The sources were found to qualify Class -43211 specifications, in accordance with AERB SS-3 and ISO-2919. (author)

Energy Technology Data Exchange (ETDEWEB)

This report provides a concise summary of the information collected and analyzed regarding the leak characteristics which define them as applicable candidates for pressure activated sealant technology. This information covers Office of Pipeline Safety reported incidents from 1985 to 1997 and was collected from existing data sources as well as operator and service company input.

Energy Technology Data Exchange (ETDEWEB)

This document is a user`s manual for the Rn3D finite element code. Rn3D was developed to simulate gas flow and radon transport in variably saturated, nonisothermal porous media. The Rn3D model is applicable to a wide range of problems involving radon transport in soil because it can simulate either steady-state or transient flow and transport in one-, two- or three-dimensions (including radially symmetric two-dimensional problems). The porous materials may be heterogeneous and anisotropic. This manual describes all pertinent mathematics related to the governing, boundary, and constitutive equations of the model, as well as the development of the finite element equations used in the code. Instructions are given for constructing Rn3D input files and executing the code, as well as a description of all output files generated by the code. Five verification problems are given that test ...

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A copyrighted computational fluid dynamics computer code, ICOMFLO, has been developed for the simulation of multiphase reacting flows. The code solves conservation equations for gaseous species and droplets (or solid particles) of various sizes. General conservation laws, expressed by elliptic type partial differential equations, are used in conjunction with rate equations governing the mass, momentum, enthalpy, species, turbulent kinetic energy, and turbulent dissipation. Associated phenomenological submodels of the code include integral combustion, two parameter turbulence, particle evaporation, and interfacial submodels. A newly developed integral combustion submodel replacing an Arrhenius type differential reaction submodel has been implemented to improve numerical convergence and enhance numerical stability. A two parameter turbulence submodel is modified for both gas and solid phases. An evaporation submodel treats ...

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FLUTAN is a highly vectorized computer code for 3D fluiddynamic and thermal-hydraulic analyses in Cartesian or cylinder coordinates. It is related to the family of COMMIX codes originally developed at Argonne National Laboratory, USA, and particularly to COMMIX-1A and COMMIX-1B, which were made available to FZK in the frame of cooperation contracts within the fast reactor safety field. FLUTAN 2.0 is an improved version of the FLUTAN code released in 1992. It offers some additional innovations, e.g. the QUICK-LECUSSO-FRAM techniques for reducing numerical diffusion in the k-{epsilon} turbulence model equations; a higher sophisticated wall model for specifying a mass flow outside the surface walls together with its flow path and its associated inlet and outlet flow temperatures; and a revised and upgraded pressure boundary condition to fully include the outlet cells in the solution process of the ...

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Grout has been evaluated as a means of containing low-level radioactive waste for long-term disposal. Concerns about hydrogen and reactive gas generation in the Hanford Site grout disposal facility lead to the development of the G3 model, which has predicted buildup of radiolytically formed gases in the grout region and their migration to air-filled regions. The G3 model is a one-dimensional code that couples models of gas diffusion with two-phase gas and liquid flow. It was coupled to a compartment diffusion model and these models predicted the concentrations of hydrogen I nitrous oxide, and other gases in air-filled regions of the grout disposal facility. The results have served as the technical basis for safety studies. A unique finding of the G3 model was that a liquid return, or expulsion of unbound liquid from the grout region, could occur. This liquid return is driven by displacement caused by gas bubbles that form within the grout`s ...

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The purpose of this paper is to compare the techniques and results obtained when a general industrial or commercial building power electrical system is prepared using the US National Electrical Code{copyright} (NEC) and the International Electrotechnical Commission (IEC) Standards. The final design offer the reader the possibility of making an initial comparison and analyzation of the two methods based of their overall effectiveness (sizing, protection, co-ordination of protective devices with components etc.), safety and cost. In order to make such a comparison a specific example was developed and limited in detail to avoid complexity in this study. The power source selected will be a delta-wye utility supplied transformer feeding various load groups such as lighting and motor circuits (possibly through associated transformers). Relevant NE Code references, IEC standards, and appropriate justifications will be made at ...

International Nuclear Information System (INIS)

The paper presents the procedure of an investigation of WWER-1000 primary circuit temperature-stress field by the use of thermohydraulic computation data for a pressurized thermal shock event ''Core overcooling''. The procedure is based on a model of the plane stress state with ideal contact between wall and medium for the calculation. The computation data are calculated on the base of WWER-1000 thermohydraulic model by the RELAP5/MOD3 codes. This model was developed jointly by the Bulgarian and BNL/USA staff to provide an analytical tool for performing safety analysis. As a result of calculations by codes the computation data for temperature field law (linear laws of a few distinguished parts) and pressure of coolant at points on inner surface of WWER-1000 primary circuit equipment are received. Such calculations can be used as a base for determination of all-important load-carrying sections of the primary circuit pipes ...

International Nuclear Information System (INIS)

The report presents the results of the study carried out to estimate the accident risk involved in the transport of spent fuel from Rajasthan Atomic Power Station near Kota to the fuel reprocessing plant at Tarapur. The technique of probabilistic safety analysis is used. The fuel considered is the Indian pressurised heavy water reactor fuel with a minimum cooling period of 485 days. The spent fuel is transported in a cuboidal, naturally-cooled shipping cask over a distance of 822 km by rail. The Indian rail accident statistics are used to estimate the basic rail accident frequency. The possible ways in which a release of radioactive material can occur from the spent fuel cask are identified by the fault tree analysis technique. The release sequences identified are classified into eight accident severity categories, and release fractions are assigned to each. The consequences resulting from the release are estimated by the computer code RADTRAN ...

International Nuclear Information System (INIS)

To resolve the central thermal safety issue for spent fuel dry storage concrete canister design or Wolsung (CANDU) nuclear power plant unit 1, a thermal analysis method has been developed for the complicated geometry of rod bundles and the multi-dimensional and multi-mode heat transfer phenomena. The canister geometry is simplified and combined heat transfer by conduction, convection, and radiation is considered through effective heat transfer coefficients. Mean temperature distributions of the fuel bundles within the fuel basket are obtained by solving the heat transfer problem using an existing computer code HEATING5. The measured steady state temperature distribution within a mock-up of a storage basket is compared to the calculated result. Steady state and/or transient fuel temperature distributions have been calculated for various ambient conditions at the canister exterior surface.

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A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

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 testing for BWR and ...

International Nuclear Information System (INIS)

Designs of nuclear reactor plants aim for high performance under safety consideration. Because of large scale and high pressure/temperature conditions, data from costly mockup tests have been required to verify simulation codes of systems and components. Establishment of design by analysis (DBA) in nuclear engineering is required for development of next generation nuclear reactors. Recent powerful computers and simulation technique enable numerical analyses to predict realistic behaviors of thermo-fluid flow, structure and do on. The present report describes resent simulation results of complex gas-liquid two-phase flow, large scale structure dynamics and fluid-structure interaction. (author)

British Library Electronic Table of Contents (United Kingdom)

To form a licensing basis for a new methodology for a fuel channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown fuel channel analysis has been developed, and tested for a high temperature thermal-chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal-chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results ...

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.

Energy Technology Data Exchange (ETDEWEB)

This paper discusses the application of a Computational Fluid Dynamics (CFD) code (FLUENT) to predict temperature and velocity profiles, and particle trajectories in an industrial fired by a low NO{sub x} pulverized coal swirl burner. The predictions of the code were compared to experimental measurements of gas temperature, particle size distribution, and particle velocities. The diagnostics employed were a suction pyrometer and a Particle Counter Sizer Velocimeter Probe (PCSV-P). Good agreement was found between the temperature predictions and the measurements. In situ particle size distributions were predicted accurately using the Baum and Street model with a swelling factor of zero. The numerical modeling indicated that the Type II flow issued from the burner could be transformed into Type III flow, by eliminating swirl from the primary air stream, which provided the preferred particle trajectories and increased the ...

International Nuclear Information System (INIS)

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

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The dogmatic and constitutional basis as well as points of definition concerning the Technical Code Clean Air (TA Luft) are explained at first, followed by an assessment of the air pollution concentrations within the overall complex of the air pollution abatement measures, and an analyses of air pollution concentrations as a concretization of the state-of-the-art. In connection with the differentiation of the various applicability levels of TA Luft, the inhouse validy for authorities, the importance for law courts and, finally, the external binding effect for plant operators and possible third parties are discussed. At the center of attention are questions concerning validity claims and validity possibilities. (HSCH).

Energy Technology Data Exchange (ETDEWEB)

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

Energy Technology Data Exchange (ETDEWEB)

Outline of three dimensional design of axial turbine was explained with examples of its application. Quasi-three dimensional design which analyzed flow in meridian plane and flow between blades seperately, and three dimensional design which, based on the quasi-three dimensional design with considering cubical force, improved distribution of static pressure and optimized distribution of repulsion and aerodynamic force, were explained. For the numerical analysis, Euler's Code was applied. AS the example of application, expectable effect by negative angle of blade was calculated, and efficiencies of single stage turbines with static blades designed by each method were compared. As the practical application of three dimensional analysis of flow to designing, Endwall Contouring for static blade of initial stage was explained. 10 refs., 12 figs.

International Nuclear Information System (INIS)

The Y2K problem may impact on nuclear installations in a number of ways because embedded systems are used in nuclear routine operation, monitoring and control system. The very simplest embedded systems are capable of performing only a single function or set of functions to meet a single predetermined purpose. In more complex systems the functioning of the embedded system is determined by an application program that enables the embedded system to be used for a particular purpose in a specific application. The simplest devices consist of a single microprocessor which may itself be packaged with other chips in a hybrid system or Application Specific Integrated Circuit (ASIC). Its input comes from a detector or sensor and its output goes to a switch or activator which may start or stop the operation of a positioning motors or, by operating a valve, may control the flow of cooling system to reactor core. Embedded systems in our ...

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.

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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.

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Sea buckthorn (Hippophae rhamnoides) is increasingly used in food supplements due to its dietary and medicinal compounds with a beneficial role in human diet and health. As many other medicinal plants, sea buckthorn can be contaminated with microorganisms which exerts an important impact on the overall quality of the products. Irradiation is an effective method for food preservation because it is able to destroy pathogenic microorganisms keeping the organoleptic and nutritional characteristics of the foods. The objective of the present study was to investigate the application of electron beam irradiation in order to improve the microbiological safety of sea buckthorn. The experimental results indicated that the electron beam treatment might be a good method to remove undesirable microorganisms from sea buckthorn without significant changes in its active principles.

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MPA research activities focus on fracture prevention and on the development of a generally applicable method of component integrity testing which, independent of the safety relevance of the components involved, is also part of ageing management. (orig.) [German] Die Forschungsaktivitaeten an der MPA im Hinblick auf die Erhoehung der Sicherheit und Zuverlaessigkeit komplex beanspruchter Bauteile orientieren sich an den Grundgedanken des Basissicherheitskonzepts (Prinzip des Bruchausschlusses). Zielsetzung war dabei die Entwicklung eines allgemein anwendbaren Nachweisverfahrens zur Komponentenintegritaet, das, abhaengig von der sicherheitstechnischen Relevanz der zu betrachtenden Komponenten, auch Bestandteil des Alterungsmanagements ist. (orig.)

British Library Electronic Table of Contents (United Kingdom)

A model for risk and reliability analysis of complex multifunctional production process systems is presented. The model employs Monte-Carlo and Markov Chain algorithms that uses a weighted index to train and simulate the fuzzy hazard data sets which represents failure outcomes of risk component transient and non-transient systems. Early simulation results shows that hazard rates and the risk of containment loss from typical floating production and storage offloading (FPSO)-Riser system for the risk components in parallel or series increases exponentially with time and decreases as safety ratings fraction increases. The reliability value decreases with time and safety fraction (SFAC) for all fuzzy hazard classifications. The results of the computed mean time before repair (MTBR) show that t...

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Metal machining facilities pose interesting design problems. This article provides an overview of key environmental control considerations and ways to deal with them. Machine shops offer a variety of environmental control challenges to mechanical engineers. Among the design problems that must be considered include: whether to provide space heating only or year-round air conditioning; temperature and relative humidity requirements; air cleanliness; exhaust air requirements; the type of HVAC system to be used; industrial waste and plumbing considerations; compressed air needs; and noise control. In addition, the designer must comply with worker health and safety and environmental protection regulations set forth by the National Institute of Safety and Health (NIOSH), OSHA, and EPA, as well as with any applicable state and local requirements.

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Liquid pipeline codes generally stipulate placement of block valves along liquid transmission pipelines such as on each side of major river crossings where environmental hazards could cause or are foreseen to potentially cause serious consequences. Codes, however, do not stipulate any requirement for block valve spacing for low vapour pressure petroleum transportation, nor for remote pipeline valve operations to reduce spills. A review of pipeline codes for valve requirement and spill limitation in high consequence areas is thus presented along with a criteria for an acceptable spill volume that could be caused by pipeline leak/full rupture. A technique for deciding economically and technically effective pipeline block valve automation for remote operation to reduce oil spilled and control of hazards is also provided. In this review, industry practice is highlighted and application of the criteria for ...

International Nuclear Information System (INIS)

Monte Carlo simulation of beam modifiers such as physical wedges and compensating filters has been performed with a rectilinear voxel geometry module. A modified version of the EGS4/DOSXYZ code has been developed for this purpose. The new implementations have been validated against the BEAM Monte Carlo code using its standard component modules (CMs) in several geometrical conditions. No significant disagreements were found within the statistical errors of 0.5% for photons and 2% for electrons. The clinical applicability and flexibility of the new version of the code has been assessed through an extensive verification versus dosimetric data. Both Varian multi-leaf collimator (MLC) wedges and standard wedges have been simulated and compared against experiments for 6 MV photon beams and different field sizes. Good agreement was found between calculated and measured depth doses and lateral dose profiles ...

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This paper describes the safety gas management system utilizing public telephone cables. Buyo Gas Co., Ltd. has installed about 100 governors for exclusive-use or use in the districts to control the gas supply to the consumers. A gas safety management system has been developed. This system consists of a pressure sensor, the terminals of gas leak alarms, and their masters. The features lie in that the system can utilize public cables, is operative at interruption of power service, can correspond to emergency situations, serves to save the cost, and can display the data constantly. For electric facilities in a governor room and the NTT cable service work, sufficient agreement with the possessor is required. In the case of non-utility electric facilities, some restrictions are imposed on the cable work for laying from privately-owned transformers, etc. Insurance for the facilities is necessary to prepare for lightning accidents. This system ...

International Nuclear Information System (INIS)

This is the first volume of a two-volume report. Volume 2 will be published at a later date. This report presents the results of a research project conducted by Brookhaven National Laboratory for the United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The purpose of the project was to develop a general methodology to be use in the assessment of the organizational factors which affect performance reliability (safety) in a nuclear power plant. The research described in this report includes the development of the Nuclear Organization and Management Analysis Concept (GNOMIC). This concept characterizes the organizational factors that impact safety performance in a nuclear power plant and identifies some methods for systematically measuring and analyzing the influence of these factors on safety performance. This report is divided into two parts; Part 1 presents an overview of the development of ...

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 ...

Energy Technology Data Exchange (ETDEWEB)

In the field of reactor and fuel cycle physics, particle transport plays and important role. Neutronic design, operation and evaluation calculations of nuclear system make use of large and powerful computer codes. However, current limitations in terms of computer resources make it necessary to introduce simplifications and approximations in order to keep calculation time and cost within reasonable limits. Two different types of methods are available in these codes. The first one is the deterministic method, which is applicable in most practical cases but requires approximations. The other method is the Monte Carlo method, which does not make these approximations but which generally requires exceedingly long running times. The main motivation of this work is to investigate the possibility of a combined use of the two methods in such a way as to retain their advantages while avoiding their drawbacks. Our work has mainly ...

International Nuclear Information System (INIS)

The safety approach of the Advanced Liquid Metal Reactor program, sponsored by the U.S. Department of Energy, relies upon passive reactor responses to off-normal conditions to limit power and temperature excursions to levels that allow large safety margins. Gas expansion modules (GEM) have been 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. Pre-application safety evaluations by the U.S. 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 positive ...

CERN Document Server

Brachytherapy is a radiotherapeutic modality that makes use of radionuclides to deliver a high radiation dose to a well-defined volume while sparing surrounding healthy structures. At the National Institute for Cancer Research of Genova a High Dose Rate remote afterloading system provides Ir(192) endocavitary brachytherapy treatments. We studied the possibility to use the Geant4 Monte Carlo simulation toolkit in brachytherapy for calculation of complex physical parameters, not directly available by experiment al measurements, used in treatment planning dose deposition models.

International Nuclear Information System (INIS)

This paper presents the results of a model for the impact of potential contamination produced by radionuclide leaking due to long-term physical, chemical and hydrogeological processes occurring in a LILW repository and in the natural environment. The analysis contains a deterministic and also a probabilistic approach for uncertainty assessment. The input function, i.e., the source term was evaluated by using AMBER code and the entire analysis was accomplished by using GOLDSIM, a powerful tool to support probabilistic simulation in management and decision-making in engineering and science. The results obtained were compared to previous simulations and uncertainty analyses (FEHM). (authors)

International Nuclear Information System (INIS)

Reduced-Moderation Water reactor (RMWR) is a light water breeder reactor developed by Japan Atomic Energy Research Institute (JAERI). The RMWR comprises tight lattice fuel assemblies with gap clearance of around 1.0 mm to reduce water volume ratio to achieve a high conversion ratio. It is important to estimate the thermal hydraulic safety margin of the tight lattice core of the RMWR. In the present study, the boiling transition (BT) prediction performance of the subchannel analysis code NASCA developed for the current BWR cores was assessed for series of tight lattice critical heat flux (CHF) experiments performed in JAERI. The test section was a 7-rod bundle with rod diameter of 12.3 mm, rod gap of 1.0 mm and heated length of 1.8m. Axial power distribution was flat. With a simple subchannel model, the code overestimates the critical power in the high mass velocity region, although the predicted critical powers in the low ...

DEFF Research Database (Denmark)

We present a new photonic technology and demonstrate that it allows for precise immobilisation of biomolecules to sensor surfaces. The technology secures spatially controlled molecular immobilisation since immobilisation of each molecule to a support surface can be limited to the focal point of the ultraviolet (UV) beam, as small as a few micrometers. We can immobilise molecules according to any pattern, from classical microarrays to diffraction patterns creating unique watermarking safety patterns. Given that suitable protein markers exists for all relevant diseases it is entirely feasible to test for a range of disease indicators (antigens and other markers) in a single test. Few micrometer spotsize allows for a virtually unlimited number of protein spots in a multipotent microarray. This new technology produces radically new photonics based microarray sensing technology and watermarking and has clear potential for biomedical, bioelectronic, surface chemistry, ...

International Nuclear Information System (INIS)

By 1970 the first edition of the ASME Code Section XI, Inservice Inspection of Nuclear Reactor Coolant Systems was published. From its inception, the Section XI inservice inspection scope was based on a fundamental risk-based selection process. In other words the inservice inspection scope included components where the consequences of a pressure boundary failure were high. Once the consequence significant system boundaries were established, inspections would then be performed at locations believed to be most susceptible service induced failure. Current Section XI requirements require that inspection locations be selected on the basis of peak stress and fatigue usage values contained in the Design Reports. These original stress calculations were designed to qualify a design and assure that the plant would provide reliable service throughout its design life. For the most part, the fatigue usage values in these reports do not provide an accurate measure of service ...

Energy Technology Data Exchange (ETDEWEB)

Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO{sub 2} fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K{sub eff} is 0.28356 well below than the critical limit, K{sub eff}=0.95 at normal condition. In a hypothetical accidental condition, the maximum K{sub eff} is found to be 0.73527 much lower than the subcritical limit. For another hypothetical ...

Energy Technology Data Exchange (ETDEWEB)

This paper is an application of the economic model developed in Part 1 and programmed in the computer code PEACES (program for the economic analysis of combined energy systems). A case study is presented in which hypothetical energy requirements at an industrial site are considered and an exercise is conducted wherein cogeneration is considered as a means of improving the energy situation at the site. Appropriate technologies that can satisfy the cogeneration requirements are investigated and technical and economic evaluations are carried out for a feasibility assessment. Of the three proposals considered, the gas turbine with heat recovery steam generator and the gas/steam turbine combined cycle cogeneration plant were found to be economically viable, while the steam turbine was not. It was recommended that the gas/steam turbine combined cycle cogeneration proposal be adopted, as it was the most economical. (Author)

International Nuclear Information System (INIS)

The buoyancy driven convective flow fields are steady circulatory flows which were made between surfaces maintained at two fixed temperatures. They are ubiquitous in nature and play an important role in many engineering applications. Especially, in last decades, natural convection in a close loop or cavity becomes the main issue in the molecular biology for the polymerase chain reaction (PCR). Application of a natural convection can reduce the costs and efforts remarkably. This paper focuses on the sensitivity study of turbulence analysis using CFD for a natural convection in a closed rectangular cavity. Using commercial CFD code, FLUENT, various turbulent models were applied to the turbulent flow. Results from each CFD model will be compared each other in the viewpoints of flow characteristics. This work will suggest the best turbulent model of CFD for analyzing turbulent flows of the natural convection in an enclosure ...

Energy Technology Data Exchange (ETDEWEB)

This paper presents the application of artificial neural networks to adiabatic flame temperature prediction of hydrocarbon fuels. The investigation was conducted over a wide range of operating conditions in terms of fuel composition, pressure and temperature of reactants, fuel-air equivalence ratio and fuel vapour fraction. Several neural network models for predicting the flame temperature for different applicable fuel ranges were built and examined. The proper preparation of network training data and the appropriate choice of network parameters for achieving better prediction accuracy are discussed. The neural network prediction results were compared with those calculated by a thermodynamic and chemical equilibrium-based computer code - the NASA program CET89. It was shown that trained neural network models can provide the adiabatic flame temperature prediction with a good level of accuracy over a wide range of operating ...

International Nuclear Information System (INIS)

The author's achievements in the title field are summarized and discussed. The following topics are dealt with: (i) principles of radionuclide X-ray fluorescence analysis; (ii) mathematical methods in X-ray fluorescence analysis; (iii) Ross differential filters; (iv) application of radionuclide X-ray fluorescence analysis in the coal industry (with emphasis on the determination of the ash content, sulfur content, and arsenic content of coal); and (v) evaluation of the X-ray fluorescence analyzer from the radiological safety point of view. (P.A.).

Energy Technology Data Exchange (ETDEWEB)

This conference was attended by experts on current issues in engineering of an urban ecosystem, composting, biofuels and green energy. The meeting provided a forum to discuss advances in the application of engineering principles and practices dealing with biological systems for the production of food, bioproducts and energy. The discussions focused on engineering practices in agriculture, forestry, bioresources, biochemistry and biosystems. The 7 technical sessions of the conference were entitled: aquaculture; safety and training; bioprocessing; energy production and biofuels; environment and ecology; soil and water; and development of technologies. The conference feature 58 presentations, of which 9 have been catalogued separately for inclusion in this database. refs., tabs., figs.

Energy Technology Data Exchange (ETDEWEB)

In this study, the state-of-the art for estimating LERF is considered for the regulatory risk-informed decisions. The consideration is mainly focused on (1) the relationship between Level 2 PSA and LERF evaluation methodology, (2) the standard requirements in terms of modeling preparation and the acceptance criteria based on the application capability II of ASME PRA standard, and (3) some pending issues for developing and proposing a simplified LERF model. This study is preliminarily presented and will be updated for establishing detailed evaluation scheme of extended MPAS (multi-purpose probabilistic analysis of safety) model and preparing the technical basis.

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)

International Nuclear Information System (INIS)

The NPPs safety system in Germany is discussed. It is shown that there exists no threat for the German NPPs at the peace times. They release insignificant quantities of radioactive substances into the water and atmosphere. The average equivalent dose constitutes 0.0005 mSv annually. The annual equivalent dose for the personnel is equal to 4.4 mSv. At the same time, the NPPs contribute to a certain degree to the environmental medium improvement, preventing the ingress therein of the sulfur and carbon dioxide, dust and nitrogen oxides by application of fossil fuels. Attention is also paid to reprocessing facilities and also to the nuclear fuel wastes disposal. The advantages of the nuclear power engineering in comparison with the fossil fuel power engineering are enumerated

International Nuclear Information System (INIS)

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.

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)

The Pacific Northwest National Laboratory (PNNL) has nearly a 40 year history of research and development in the field of nondestructive evaluation (NDE). One area of NDE expertise at PNNL is electromagnetic testing which includes a field of eddy current testing (ET). One benefit is that ET can typically be performed at high speeds, and as a result has found many applications in process monitoring and poduction lines. ET has been used in the nuclear, aerospace, and automotive industries for many years. Et technology lends itself well to the detection of near-surface or surface breaking defects such as surface scratches. This paper provides an overview of theory regarding the usage of ET, selected application studies performed by PNNL, a safety analysis, and a wrtie up pertaining to the operations of ET to detect surface scratches.

Energy Technology Data Exchange (ETDEWEB)

A study has been conducted on the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. Studies were conducted on three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparisons, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change.

Energy Technology Data Exchange (ETDEWEB)

In order to study nuclear criticality safety related to the development of fast breeder technology, /sup 252/Cf-source-driven neutron noise analysis measurements were performed with a Pu-U nitrate solution in a slab tank of various heights and thickness varying 11.43 cm to 19.05 cm. The results and conclusions of these experiments are (1) a capability to measure the subcriticality of a multiplying system of slab geometry to a k/sub eff/ as low as 0.7 was demonstrated, (2) calculated neutron multiplication factors agreed with those from the experiments within approx.0.02, and (3) the applicability of the method for plutonium solution systems was demonstrated. This paper describes measurements in which the height of the slab was varied for a fixed thickness and the thickness varied for a fixed height, which are the first applications of this measurement method to slab geometry.

International Nuclear Information System (INIS)

Purpose: The purpose of this course is to review the physical principles underlying design, clinical application and execution of interstitial and intracavitary implants in the classical low dose-rate (LDR) range. This year, the course will focus on quality assurance of sources, applicators and treatment planning software. In addition, development of procedures and QA checks designed optimize treatment delivery accuracy and patient safety during each individual procedure will be reviewed. The level of presentation will be designed to accommodate both physicists and physicians. Implementation of recently published AAPM Task Group reports (no. 40, 'Comprehensive Quality Assurance' and No. 43, 'Dosimetry of Interstitial Brachytherapy Sources') will be reviewed. Outline: (A) General Principles (1) QA endpoints: temporal accuracy, positional accuracy, dose delivery accuracy, and safety of the patient, ...

Energy Technology Data Exchange (ETDEWEB)

The Rocky Flats Plant, located approximately 26 km northwest of downtown Denver, Colorado, has developed an emergency response atmospheric dispersion model for complex terrain applications. Plant personnel would use the model, known as the Terrain-Responsive Atmospheric Code (TRAC) (Hodgin 1985) to project plume impacts and provide off-site protective action recommendations to the State of Colorado should a hazardous material release occur from the facility. The Colorado Department of Health (CDH) entered into an interagency agreement with the Rocky Flats Plant prime contractor, EG&G Rocky Flats, and the US Department of Energy to evaluate TRAC as an acceptable emergency response tool. After exhaustive research of similar evaluation processes from other emergency response and regulatory organizations, the interagency committee devised a formal acceptance process. The process contains an evaluation protocol (Hodgin and Smith 1992), ...

Energy Technology Data Exchange (ETDEWEB)

Safety in general and harmonization of International Maritime Organization rules on mobile offshore drilling rig operation in particular are discussed. The improvement of the industry's safety record is also discussed.

Science.gov (United States)

PMID:7561458

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) ...

International Nuclear Information System (INIS)

The world's fleet of nuclear power plants is, on average, more than 20 years old. Even though the design life of a nuclear power plant is typically 30-40 years, it is quite feasible that many plants will be able to operate in excess of their design lives, provided that nuclear power plant engineers demonstrate by analysis, trending, equipment and system upgrades, increased vigilance, testing and ageing management that the plant will operate safely. In the operation of nuclear power plants, safety should be always the prime consideration. Plant operators and regulators must always ensure that plant safety is maintained, and where possible enhanced, during a plant's operating lifetime. Nuclear power plant life management (PLiM) has gained increased attention over the past decade, and effective ageing management of systems, structures and components (SSCs) is a key element in PLiM for the safe and reliable long term operation of nuclear power ...

Science.gov (United States)

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give complete insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with ...

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 cylinder and dome surface area ratios than were available in the ...

Energy Technology Data Exchange (ETDEWEB)

Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. The cases of GDH impact on an adjacent GDH and its attached piping are investigated in this paper. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would ...

International Nuclear Information System (INIS)

The O(#alpha#) electroweak radiative corrections to #gamma##gamma##->#WW#->#4f within the electroweak standard model are calculated in double-pole approximation (DPA). Virtual corrections are treated in DPA, leading to a classification into factorizable and non-factorizable contributions, and real-photonic corrections are based on complete lowest-order matrix elements for #gamma##gamma##->#4f+#gamma#. Soft and collinear singularities appearing in the virtual and real corrections are combined alternatively in two different ways, namely by using the dipole subtraction method or by applying phase-space slicing. The radiative corrections are implemented in a Monte Carlo generator called Coffer #gamma##gamma# - the computer code can be obtained from the authors upon request - which optionally includes anomalous triple and quartic gauge-boson couplings in addition and performs a convolution over realistic spectra of the photon beams. A detailed survey of ...

International Nuclear Information System (INIS)

Full text of publication follows: Hydrogen release and distribution in nuclear power plant containment is an important safety issue. Selection of a proper turbulence model is important for accurate estimation of the mixing process. The selection of turbulence model is dictated by the best compromise between accuracy and computational efforts. For this, three different turbulence models, viz. Standard k-#epsilon#, RNG k-#epsilon# and Reynolds Stress Model, based on Reynolds averaged Navier Stokes equations (RANS) approach, were used. The computations were done using the CFD code FLUENT, which is based on the control volume methodology. The computational results were compared with the experimental results of HYMIS test facility, where helium was used to simulate hydrogen. The processes of helium plume rise, multiple plume merging, distribution and mixing were studied. Based on these computations, a simple analytical/empirical zone based model was ...

International Nuclear Information System (INIS)

Total quality management strategy and the worldwide introduction of the DIN/ISO 9000 (EN 29 000) series of standards have given new impetus to traditional quality assurance. The most important change must surely be seen in the holistic approach of total quality management and its strict orientation towards customer requirements and satisfaction. International codes and standards for the nuclear industry will also have to be brought into line as part of the process of harmonizing quality assurance system standards. One possible approach is simply to specify a supplementary 'delta' of nuclear-specific requirements to be appended to the broad range of conventional requirements. It is a particular feature of quality-assured procedures in Germany that product and/or component related quality requirements and quality verifications are defined in the specifications of the architect engineer so that full implementation of the requirements from the design phase through to ...

International Nuclear Information System (INIS)

The present emphasis on inherent safety for liquid-metal reactor designs has resulted in a need to represent the various reactivity feedback mechanisms as accurately as possible. The dominant negative reactivity feedback has been found to result from radial expansion of the core for most postulated anticipated transient without scram events. For this reasons, a more detailed model for calculating the reactivity feedback from radial core expansion, including subassembly bowing has been recently developed for use with the SASSYS/SAS4A code system. The purpose of this summary is to present an extension to the model so that it is more suitable for handling a core restraint design as used in the Fast Flux Test Facility (FFTF), and to compare the SASSYS/SAS4A results using this model to the empirical correlation presently being used to account for radial core expansion reactivity feedback in the FFTF.

Energy Technology Data Exchange (ETDEWEB)

Los Alamos National Laboratory is subject to annual emissions-reporting requirements for regulated air contaminants under Title 20 of the New Mexico Administrative Code, Chapter 2, Part 73 (20.2.73 NMAC), Notice of Intent and Emissions Inventory Requirements. The applicability of the requirements is based on the Laboratory's potential to emit 100 tons per year of suspended particulate matter, nitrogen oxides, carbon monoxide, sulfur oxides, or volatile organic compounds. For calendar year 2001, the Technical Area 3 steam plant was the primary source of criteria air pollutants from the Laboratory, while research and development activities were the primary source of volatile organic compounds. Emissions of beryllium and aluminum were reported for activities permitted under 20.2.72 NMAC. Hazardous air pollutant emissions from chemical use for research and development activities were also reported.

CERN Document Server

Physical-layer network coding (PNC) is a promising approach for wireless networks. It allows nodes to transmit simultaneously. Due to the difficulties of scheduling simultaneous transmissions, existing works on PNC are based on simplified medium access control (MAC) protocols, which are not applicable to general multi-hop wireless networks, to the best of our knowledge. In this paper, we propose a distributed MAC protocol that supports PNC in multi-hop wireless networks. The proposed MAC protocol is based on the carrier sense multiple access (CSMA) strategy and can be regarded as an extension to the IEEE 802.11 MAC protocol. In the proposed protocol, each node collects information on the queue status of its neighboring nodes. When a node finds that there is an opportunity for some of its neighbors to perform PNC, it notifies its corresponding neighboring nodes and initiates the process of packet exchange using PNC, with the node itself as a ...

Energy Technology Data Exchange (ETDEWEB)

Real-time mission-oriented embedded systems are much more difficult to design than ordinary software systems. They require highly reliable and efficient implementations to satisfy mission and time constraints imposed by the applications. The Ada language was designed to facilitate real-time-system software development. However, for many programmers the size and complexity of Ada itself are of concern. In the assertive programming paradigm, computations are specified as sets of assertions about properties of the solution, and not as a sequence of procedural steps. Solving procedures are automatically generated from the assertive description. Real-time programming for mission-oriented systems is supported by equational languages in which assertions are expressed as algebraic equations. Programs written in equational languages are concise, free from implementation details, and easily amenable to verification and parallel processing. The level of programming expertise ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

Located approximately 350 km east of Newfoundland, on the eastern edge of the Jeanne d'Arc Basin, and approximately 50 km from both the Terra Nova and Hibernia oilfields, the White Rose oilfield is considered the main oil producing basin off the eastern coast of North America. Husky Oil is very active in the region and holds almost 32 per cent net working interest in the Significant Discovery License areas in the Jeanne d'Arc basin. The development of an economically significant oil discovery in the White Rose Significant Discovery Area is being proposed jointly by Husky Oil and Petro-Canada. It is estimated that the project would enable the recovery of an estimated 36 million cubic meters (230 million barrels) of recoverable oil in this area that covers approximately 40 square kilometers. The operators propose to use a ship-shaped floating production, storage and offloading (FPSO) facility which could store up to 135,000 cubic meters (850,000 barrels) of oil or the ...

KoreaScience (Korea)

Full Text Available

Science.gov (United States)

PMID:19955450

Science.gov (United States)

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Energy Technology Data Exchange (ETDEWEB)

Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R&D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. ...

Energy Technology Data Exchange (ETDEWEB)

This project aims at developing risk-informed application technologies to enhance the safety and economy of nuclear power plant altogether. For this, the Integrated Level 1 and 2 PSA model is developed. In addition, the fire and internal flooding PSA models are improved according to the PSA standard of U.S.A. To solve the issues of domestic PSA model, the best-estimate thermal hydraulic analyses are preformed for the ATWS and LSSB. In order to reduce the uncertainty of PSA, several new PSA technologies are developed: (1) more exact quantification of large fault tree, (2) importance measure including the effects of external PSA. As feasibility studies of Option 2 and 3, the class of 6 systems' SSC are re-classified based on the risk information and the sensitivity analyses is performed for the EDG starting time, respectively. It is also improved that the methodology to identify the vital area of NPP. The research results of this project ...

International Nuclear Information System (INIS)

The Advanced CANDU Reactor (ACR"T"M) is the 'Next Generation' CANDU"R reactor, aimed at safe, reliable power production at a capital cost significantly less than that of current reactors such as the very successful CANDU 6 reactors (e.g., Wolsong 1-4). The Wolsong 1-4 units are being joined by the Qinshan Phase 3 units in China as the current successful examples of CANDU technology. The ACR design builds on this knowledge base, adding a selected group of innovations to obtain substantial cost reduction while retaining a proven design basis. The ACR maximizes the use of components and equipment applications that are well proven through CANDU and other nuclear experience. Joint development of equipment designs and specifications with manufactures has been emphasized. Similarly, the ACR design emphasizes constructability, and takes advantage of inherent CANDU features to enable short project and construction schedules. Overall, the ACR design represents a balance of ...

International Nuclear Information System (INIS)

This document presents the way which the 8th principle of the quality management system 'Process approach' is applied, the principle that is identified and used by international standard ISO 9000. In order to understand the evolution of the management system requirements, as used today in different activities namely, industry, services, and nuclear activities, the authors present an evolution of the quality concept and its traceability to different standards, applicable in time. There are described the requirements of ISO 9001 standard, that represents the most widely spread model for modern organization management and the IAEA concerns related to integration of the above standard requirements into the most recent safety IAEA standard 'The Management System for facilities and activities'. The IAEA Safety Standard GS-R-3 describes a management model considering both the evolution of the quality requirements into the modern ...

International Nuclear Information System (INIS)

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 software cost reduction project of the U.S. naval research laboratory, and is, as a whole, clear and well structured except for the modules related to the operation of the computer itself. These ...

International Nuclear Information System (INIS)

The DOE is conducting a comprehensive technical analysis of a flexible-fuel transportation system in the United States -- that is, a system that could easily switch between petroleum and another fuel, depending on price and availability. The DOE Alternative Fuels Assessment is aimed directly at questions of energy security and fuel availability, but covers a wide range of issues. This report examines environmental, health, and safety concerns associated with a switch to alternative- and flexible-fuel vehicles. Three potential alternatives to oil-based fuels in the transportation sector are considered: methanol, compressed natural gas (CNG), and electricity. The objective is to describe and discuss qualitatively potential environmental, health, and safety issues that would accompany widespread use of these three fuels. This report presents the results of exhaustive literature reviews; discussions with specialists in the vehicular and ...

Energy Technology Data Exchange (ETDEWEB)

On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad. Shortly after the accident occurred, the USNRC met with the National Transportation Safety Board (NTSB), the U.S. agency responsible for determining the cause of transportation accidents, to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory ...

International Nuclear Information System (INIS)

In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, proposal for ASTEC ...

International Nuclear Information System (INIS)

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

Energy Technology Data Exchange (ETDEWEB)

The general strategy for improving the safety of nuclear power plant and its economics is to accomplish power uprates while securing sufficient thermalhydraulic margin. In order to succeed this strategy, there have been a lot of efforts in increasing the margin through the enhancement of heat transfer capability in coolants. However, despite their efforts, only about 10 {approx} 15 % increase of the thermal margin is possible by using the best art known well up to now with installation of mechanical engineering devices such as mixing vane or button to generating the swirl flow and turbulent mixing. The limit of the capability of the best technique has made a lot of engineers to be frustrated to do the power uprates. Nevertheless, fortunately a new innovative idea is being proposed in heat transfer community as an engineering colloidal fluid to basically change the original properties of the coolant. The fluid began to be called by Choi as a nanofluid which is a ...

International Nuclear Information System (INIS)

A number of test series to investigate passive safety systems for the next generation of Light Water Reactors have been performed in the PANDA multi-purpose facility at the Paul Scherrer Institut (PSI). The large scale thermal-hydraulic test facility allows to investigate LWR containment phenomena and system behaviour. PANDA was first used to examine the Passive Containment Cooling System (PCCS) for the Simplified Boiling Water Reactor (SBWR). In 1996 new test series were initiated; all related to projects of the EC Fourth Framework Programme on Nuclear Fission Safety. One of these projects (TEPSS) is focused on the European Simplified Boiling Water Reactor (ESBWR). The ESBWR containment features and PCCS long-term post LOCA response were investigated in PANDA. The PCCS start-up was demonstrated, the effect of nitrogen hidden somewhere in the drywell and released later in the transient was simulated and the effect of light gases (helium) on the ...