A Formal Verification Method of Function Block Diagram

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun; Jee, Eun Kyoung; Jeon, Seung Jae; Park, Gee Yong; Kwon, Kee Choon

2007-01-01

Programmable Logic Controller (PLC), an industrial computer specialized for real-time applications, is widely used in diverse control systems in chemical processing plants, nuclear power plants or traffic control systems. As a PLC is often used to implement safety, critical embedded software, rigorous safety demonstration of PLC code is necessary. Function block diagram (FBD) is a standard application programming language for the PLC and currently being used in the development of a fully-digitalized reactor protection system (RPS), which is called the IDiPS, under the KNICS project. Therefore, verification issue of FBD programs is a pressing problem, and hence is of great importance. In this paper, we propose a formal verification method of FBD programs; we defined FBD programs formally in compliance with IEC 61131-3, and then translate the programs into Verilog model, and finally the model is verified using a model checker SMV. To demonstrate the feasibility and effective of this approach, we applied it to IDiPS which currently being developed under KNICS project. The remainder of this paper is organized as follows. Section 2 briefly describes Verilog and Cadence SMV. In Section 3, we introduce FBD2V which is a tool implemented to support the proposed FBD verification framework. A summary and conclusion are provided in Section 4

Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

International Nuclear Information System (INIS)

Ruebel, Andre; Meleshyn, Artur

2014-08-01

The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

Verification of chemistry reference ranges using a simple method in sub-Saharan Africa.

Science.gov (United States)

De Baetselier, Irith; Taylor, Douglas; Mandala, Justin; Nanda, Kavita; Van Campenhout, Christel; Agingu, Walter; Madurai, Lorna; Barsch, Eva-Maria; Deese, Jennifer; Van Damme, Lut; Crucitti, Tania

2016-01-01

Chemistry safety assessments are interpreted by using chemistry reference ranges (CRRs). Verification of CRRs is time consuming and often requires a statistical background. We report on an easy and cost-saving method to verify CRRs. Using a former method introduced by Sigma Diagnostics, three study sites in sub-Saharan Africa, Bondo, Kenya, and Pretoria and Bloemfontein, South Africa, verified the CRRs for hepatic and renal biochemistry assays performed during a clinical trial of HIV antiretroviral pre-exposure prophylaxis. The aspartate aminotransferase/alanine aminotransferase, creatinine and phosphorus results from 10 clinically-healthy participants at the screening visit were used. In the event the CRRs did not pass the verification, new CRRs had to be calculated based on 40 clinically-healthy participants. Within a few weeks, the study sites accomplished verification of the CRRs without additional costs. The aspartate aminotransferase reference ranges for the Bondo, Kenya site and the alanine aminotransferase reference ranges for the Pretoria, South Africa site required adjustment. The phosphorus CRR passed verification and the creatinine CRR required adjustment at every site. The newly-established CRR intervals were narrower than the CRRs used previously at these study sites due to decreases in the upper limits of the reference ranges. As a result, more toxicities were detected. To ensure the safety of clinical trial participants, verification of CRRs should be standard practice in clinical trials conducted in settings where the CRR has not been validated for the local population. This verification method is simple, inexpensive, and can be performed by any medical laboratory.

Methods and practices for verification and validation of programmable systems

International Nuclear Information System (INIS)

Heimbuerger, H.; Haapanen, P.; Pulkkinen, U.

1993-01-01

The programmable systems deviate by their properties and behaviour from the conventional non-programmable systems in such extent, that their verification and validation for safety critical applications requires new methods and practices. The safety assessment can not be based on conventional probabilistic methods due to the difficulties in the quantification of the reliability of the software and hardware. The reliability estimate of the system must be based on qualitative arguments linked to a conservative claim limit. Due to the uncertainty of the quantitative reliability estimate other means must be used to get more assurance about the system safety. Methods and practices based on research done by VTT for STUK, are discussed in the paper as well as the methods applicable in the reliability analysis of software based safety functions. The most essential concepts and models of quantitative reliability analysis are described. The application of software models in probabilistic safety analysis (PSA) is evaluated. (author). 18 refs

VERIFICATION OF THE FOOD SAFETY MANAGEMENT SYSTEM IN DEEP FROZEN FOOD PRODUCTION PLANT

Directory of Open Access Journals (Sweden)

Peter Zajác

2010-07-01

Full Text Available In work is presented verification of food safety management system of deep frozen food. Main emphasis is on creating set of verification questions within articles of standard STN EN ISO 22000:2006 and on searching of effectiveness in food safety management system. Information were acquired from scientific literature sources and they pointed out importance of implementation and upkeep of effective food safety management system. doi:10.5219/28

Verification of the safety communication protocol in train control system using colored Petri net

International Nuclear Information System (INIS)

Chen Lijie; Tang Tao; Zhao Xianqiong; Schnieder, Eckehard

2012-01-01

This paper deals with formal and simulation-based verification of the safety communication protocol in ETCS (European Train Control System). The safety communication protocol controls the establishment of safety connection between train and trackside. Because of its graphical user interface and modeling flexibility upon the changes in the system conditions, this paper proposes a composition Colored Petri Net (CPN) representation for both the logic and the timed model. The logic of the protocol is proved to be safe by means of state space analysis: the dead markings are correct; there are no dead transitions; being fair. Further analysis results have been obtained using formal and simulation-based verification approach. The timed models for the open transmit system and the application process are created for the purpose of performance analysis of the safety communication protocol. The models describe the procedure of data transmission and processing, and also provide relevant timed and stochastic factors, as well as time delay and lost packet, which may influence the time for establishment of safety connection of the protocol. Time for establishment of safety connection of the protocol in normal state is verified by formal verification, and then time for establishment of safety connection with different probability of lost packet is simulated. After verification it is found that the time for establishment of safety connection of the safety communication protocol satisfies the safety requirements.

SCALE criticality safety verification and validation package

International Nuclear Information System (INIS)

Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

1998-01-01

Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

Improved verification methods for safeguards verifications at enrichment plants

International Nuclear Information System (INIS)

Lebrun, A.; Kane, S. C.; Bourva, L.; Poirier, S.; Loghin, N. E.; Langlands, D.

2009-01-01

The International Atomic Energy Agency (IAEA) has initiated a coordinated research and development programme to improve its verification methods and equipment applicable to enrichment plants. The programme entails several individual projects to meet the objectives of the IAEA Safeguards Model Approach for Gas Centrifuge Enrichment Plants updated in 2006. Upgrades of verification methods to confirm the absence of HEU (highly enriched uranium) production have been initiated and, in particular, the Cascade Header Enrichment Monitor (CHEM) has been redesigned to reduce its weight and incorporate an electrically cooled germanium detector. Such detectors are also introduced to improve the attended verification of UF 6 cylinders for the verification of the material balance. Data sharing of authenticated operator weighing systems such as accountancy scales and process load cells is also investigated as a cost efficient and an effective safeguards measure combined with unannounced inspections, surveillance and non-destructive assay (NDA) measurement. (authors)

Improved verification methods for safeguards verifications at enrichment plants

Energy Technology Data Exchange (ETDEWEB)

Lebrun, A.; Kane, S. C.; Bourva, L.; Poirier, S.; Loghin, N. E.; Langlands, D. [Department of Safeguards, International Atomic Energy Agency, Wagramer Strasse 5, A1400 Vienna (Austria)

2009-07-01

The International Atomic Energy Agency (IAEA) has initiated a coordinated research and development programme to improve its verification methods and equipment applicable to enrichment plants. The programme entails several individual projects to meet the objectives of the IAEA Safeguards Model Approach for Gas Centrifuge Enrichment Plants updated in 2006. Upgrades of verification methods to confirm the absence of HEU (highly enriched uranium) production have been initiated and, in particular, the Cascade Header Enrichment Monitor (CHEM) has been redesigned to reduce its weight and incorporate an electrically cooled germanium detector. Such detectors are also introduced to improve the attended verification of UF{sub 6} cylinders for the verification of the material balance. Data sharing of authenticated operator weighing systems such as accountancy scales and process load cells is also investigated as a cost efficient and an effective safeguards measure combined with unannounced inspections, surveillance and non-destructive assay (NDA) measurement. (authors)

78 FR 32010 - Pipeline Safety: Public Workshop on Integrity Verification Process

Science.gov (United States)

2013-05-28

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No... Hazardous Materials Safety Administration, DOT. ACTION: Notice of public meeting. SUMMARY: This notice is announcing a public workshop to be held on the concept of ``Integrity Verification Process.'' The Integrity...

Verification of MCNP6.2 for Nuclear Criticality Safety Applications

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-05-10

Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.

Towards the Verification of Safety-critical Autonomous Systems in Dynamic Environments

Directory of Open Access Journals (Sweden)

Adina Aniculaesei

2016-12-01

Full Text Available There is an increasing necessity to deploy autonomous systems in highly heterogeneous, dynamic environments, e.g. service robots in hospitals or autonomous cars on highways. Due to the uncertainty in these environments, the verification results obtained with respect to the system and environment models at design-time might not be transferable to the system behavior at run time. For autonomous systems operating in dynamic environments, safety of motion and collision avoidance are critical requirements. With regard to these requirements, Macek et al. [6] define the passive safety property, which requires that no collision can occur while the autonomous system is moving. To verify this property, we adopt a two phase process which combines static verification methods, used at design time, with dynamic ones, used at run time. In the design phase, we exploit UPPAAL to formalize the autonomous system and its environment as timed automata and the safety property as TCTL formula and to verify the correctness of these models with respect to this property. For the runtime phase, we build a monitor to check whether the assumptions made at design time are also correct at run time. If the current system observations of the environment do not correspond to the initial system assumptions, the monitor sends feedback to the system and the system enters a passive safe state.

Motion simulation of hydraulic driven safety rod using FSI method

International Nuclear Information System (INIS)

Jung, Jaeho; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

2013-01-01

Hydraulic driven safety rod which is one of them is being developed by Division for Reactor Mechanical Engineering, KAERI. In this paper the motion of this rod is simulated by fluid structure interaction (FSI) method before manufacturing for design verification and pump sizing. A newly designed hydraulic driven safety rod which is one of reactivity control mechanism is simulated using FSI method for design verification and pump sizing. The simulation is done in CFD domain with UDF. The pressure drop is changed slightly by flow rates. It means that the pressure drop is mainly determined by weight of moving part. The simulated velocity of piston is linearly proportional to flow rates so the pump can be sized easily according to the rising and drop time requirement of the safety rod using the simulation results

Methods of Software Verification

Directory of Open Access Journals (Sweden)

R. E. Gurin

2015-01-01

Full Text Available This article is devoted to the problem of software verification (SW. Methods of software verification designed to check the software for compliance with the stated requirements such as correctness, system security and system adaptability to small changes in the environment, portability and compatibility, etc. These are various methods both by the operation process and by the way of achieving result. The article describes the static and dynamic methods of software verification and paid attention to the method of symbolic execution. In its review of static analysis are discussed and described the deductive method, and methods for testing the model. A relevant issue of the pros and cons of a particular method is emphasized. The article considers classification of test techniques for each method. In this paper we present and analyze the characteristics and mechanisms of the static analysis of dependencies, as well as their views, which can reduce the number of false positives in situations where the current state of the program combines two or more states obtained both in different paths of execution and in working with multiple object values. Dependences connect various types of software objects: single variables, the elements of composite variables (structure fields, array elements, the size of the heap areas, the length of lines, the number of initialized array elements in the verification code using static methods. The article pays attention to the identification of dependencies within the framework of the abstract interpretation, as well as gives an overview and analysis of the inference tools.Methods of dynamic analysis such as testing, monitoring and profiling are presented and analyzed. Also some kinds of tools are considered which can be applied to the software when using the methods of dynamic analysis. Based on the work a conclusion is drawn, which describes the most relevant problems of analysis techniques, methods of their solutions and

A Survey on Formal Verification Techniques for Safety-Critical Systems-on-Chip

Directory of Open Access Journals (Sweden)

Tomás Grimm

2018-05-01

Full Text Available The high degree of miniaturization in the electronics industry has been, for several years, a driver to push embedded systems to different fields and applications. One example is safety-critical systems, where the compactness in the form factor helps to reduce the costs and allows for the implementation of new techniques. The automotive industry is a great example of a safety-critical area with a great rise in the adoption of microelectronics. With it came the creation of the ISO 26262 standard with the goal of guaranteeing a high level of dependability in the designs. Other areas in the safety-critical applications domain have similar standards. However, these standards are mostly guidelines to make sure that designs reach the desired dependability level without explicit instructions. In the end, the success of the design to fulfill the standard is the result of a thorough verification process. Naturally, the goal of any verification team dealing with such important designs is complete coverage as well as standards conformity, but as these are complex hardware, complete functional verification is a difficult task. From the several techniques that exist to verify hardware, where each has its pros and cons, we studied six well-established in academia and in industry. We can divide them into two categories: simulation, which needs extremely large amounts of time, and formal verification, which needs unrealistic amounts of resources. Therefore, we conclude that a hybrid approach offers the best balance between simulation (time and formal verification (resources.

Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 & Vol 2

Energy Technology Data Exchange (ETDEWEB)

PARSONS, J.E.

2000-07-15

The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented.

Probabilistic Requirements (Partial) Verification Methods Best Practices Improvement. Variables Acceptance Sampling Calculators: Derivations and Verification of Plans. Volume 1

Science.gov (United States)

Johnson, Kenneth L.; White, K, Preston, Jr.

2012-01-01

The NASA Engineering and Safety Center was requested to improve on the Best Practices document produced for the NESC assessment, Verification of Probabilistic Requirements for the Constellation Program, by giving a recommended procedure for using acceptance sampling by variables techniques. This recommended procedure would be used as an alternative to the potentially resource-intensive acceptance sampling by attributes method given in the document. This document contains the outcome of the assessment.

Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 and Vol 2

CERN Document Server

Parsons, J E

2000-01-01

The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented.

Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 and Vol 2

International Nuclear Information System (INIS)

PARSONS, J.E.

2000-01-01

The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented

Response to "Improving Patient Safety With Error Identification in Chemotherapy Orders by Verification Nurses"
.

Science.gov (United States)

Zhu, Ling-Ling; Lv, Na; Zhou, Quan

2016-12-01

We read, with great interest, the study by Baldwin and Rodriguez (2016), which described the role of the verification nurse and details the verification process in identifying errors related to chemotherapy orders. We strongly agree with their findings that a verification nurse, collaborating closely with the prescribing physician, pharmacist, and treating nurse, can better identify errors and maintain safety during chemotherapy administration.

HTGR analytical methods and design verification

International Nuclear Information System (INIS)

Neylan, A.J.; Northup, T.E.

1982-05-01

Analytical methods for the high-temperature gas-cooled reactor (HTGR) include development, update, verification, documentation, and maintenance of all computer codes for HTGR design and analysis. This paper presents selected nuclear, structural mechanics, seismic, and systems analytical methods related to the HTGR core. This paper also reviews design verification tests in the reactor core, reactor internals, steam generator, and thermal barrier

Safety verification of non-linear hybrid systems is quasi-decidable

Czech Academy of Sciences Publication Activity Database

Ratschan, Stefan

2014-01-01

Roč. 44, č. 1 (2014), s. 71-90 ISSN 0925-9856 R&D Projects: GA ČR GCP202/12/J060 Institutional support: RVO:67985807 Keywords : hybrid system s * safety verification * decidability * robustness Subject RIV: IN - Informatics, Computer Science Impact factor: 0.875, year: 2014

Spent fuel verification options for final repository safeguards in Finland. A study on verification methods, their feasibility and safety aspects

International Nuclear Information System (INIS)

Hautamaeki, J.; Tiitta, A.

2000-12-01

The verification possibilities of the spent fuel assemblies from the Olkiluoto and Loviisa NPPs and the fuel rods from the research reactor of VTT are contemplated in this report. The spent fuel assemblies have to be verified at the partial defect level before the final disposal into the geologic repository. The rods from the research reactor may be verified at the gross defect level. Developing a measurement system for partial defect verification is a complicated and time-consuming task. The Passive High Energy Gamma Emission Tomography and the Fork Detector combined with Gamma Spectrometry are the most potential measurement principles to be developed for this purpose. The whole verification process has to be planned to be as slick as possible. An early start in the planning of the verification and developing the measurement devices is important in order to enable a smooth integration of the verification measurements into the conditioning and disposal process. The IAEA and Euratom have not yet concluded the safeguards criteria for the final disposal. E.g. criteria connected to the selection of the best place to perform the verification. Measurements have not yet been concluded. Options for the verification places have been considered in this report. One option for a verification measurement place is the intermediate storage. The other option is the encapsulation plant. Crucial viewpoints are such as which one offers the best practical possibilities to perform the measurements effectively and which would be the better place in the safeguards point of view. Verification measurements may be needed both in the intermediate storages and in the encapsulation plant. In this report also the integrity of the fuel assemblies after wet intermediate storage period is assessed, because the assemblies have to stand the handling operations of the verification measurements. (orig.)

The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

International Nuclear Information System (INIS)

Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

2007-01-01

The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods

Co-simulation for real time safety verification of nuclear power plants

International Nuclear Information System (INIS)

Boafo, E.K.; Zhang, L.; Nasimi, E.; Gabbar, H.A.

2015-01-01

Small and major accidents and near misses are still occurring in nuclear power plants (NPPs). Risk level has increased with the degradation of NPP equipment and instrumentations. In order to achieve NPP safety, it is important to continuously evaluate risk for all potential hazard and fault propagation scenarios and map protection layers to fault / failure / hazard propagation scenarios to be able to evaluate and verify safety level during NPP operation. There are major limitations in current real time safety verification tools, as it is mainly offline and with no integration to NPP simulation tools. The main goal of this research is to develop real time safety verification with co-simulation tool to be integrated with plant operation support systems. This includes the development of static and dynamic fault semantic network (FSN) to model all possible fault propagation scenarios and the interrelationships among associated process variables. Safety and protection layers along with their reliability are mapped to FSN so that safety levels can be verified during plant operation. Errors between multiphysics models and real time data are modeled to accurately and dynamically tune FSN for each fault propagation scenario. The detailed methodology will show how to integrate process models, construction of static FSN with fault propagation scenarios, and evaluation and tuning of dynamic FSN with probabilistic and process variable interaction values. Principle Component Analysis method is used reduce dimensionality and reduce process variables associated with each fault scenario. Then map independent protection layers (IPL) to FSN with estimated reliability measures of each protection layer to accurately verify safety for different operational scenarios. Intelligent algorithms is used with multivariate techniques to accurate define the interrelation among process variables, in terms of signal strength and time delay, using Genetic Programming (GP), which will provide basis

Seismic verification methods for structures and equipment of VVER-type and RBMK-type NPPs (summary of experiences)

International Nuclear Information System (INIS)

Masopust, R.

2003-01-01

The main verification methods for structures and equipment of already existing VVER-type and RBMK-type NPPs are briefly described. The following aspects are discussed: fundamental seismic safety assessment principles for VVER/RBMK-type NPPs (seismic safety assessment procedure, typical work plan for seismic safety assessment of existing NPPs, SMA (HCLPF) calculations, modified GIP (GIP-VVER) procedure, similarity of VVER/RBMK equipment to that included in the SQUG databases and seismic interactions

Application of Software Safety Analysis Methods

International Nuclear Information System (INIS)

Park, G. Y.; Hur, S.; Cheon, S. W.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.; Lee, S. J.; Koo, Y. H.

2009-01-01

A fully digitalized reactor protection system, which is called the IDiPS-RPS, was developed through the KNICS project. The IDiPS-RPS has four redundant and separated channels. Each channel is mainly composed of a group of bistable processors which redundantly compare process variables with their corresponding setpoints and a group of coincidence processors that generate a final trip signal when a trip condition is satisfied. Each channel also contains a test processor called the ATIP and a display and command processor called the COM. All the functions were implemented in software. During the development of the safety software, various software safety analysis methods were applied, in parallel to the verification and validation (V and V) activities, along the software development life cycle. The software safety analysis methods employed were the software hazard and operability (Software HAZOP) study, the software fault tree analysis (Software FTA), and the software failure modes and effects analysis (Software FMEA)

Verification and validation benchmarks.

Energy Technology Data Exchange (ETDEWEB)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-02-01

Verification and validation (V&V) are the primary means to assess the accuracy and reliability of computational simulations. V&V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V&V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the level of

Functional verification of a safety class controller for NPPs using a UVM register Model

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyu Chull [Dept. of Applied Computer Engineering, Dankook University, Cheonan (Korea, Republic of)

2014-06-15

A highly reliable safety class controller for NPPs (Nuclear Power Plants) is mandatory as even a minor malfunction can lead to disastrous consequences for people, the environment or the facility. In order to enhance the reliability of a safety class digital controller for NPPs, we employed a diversity approach, in which a PLC-type controller and a PLD-type controller are to be operated in parallel. We built and used structured testbenches based on the classes supported by UVM for functional verification of the PLD-type controller designed for NPPs. We incorporated a UVM register model into the testbenches in order to increase the controllability and the observability of the DUT(Device Under Test). With the increased testability, we could easily verify the datapaths between I/O ports and the register sets of the DUT, otherwise we had to perform black box tests for the datapaths, which is very cumbersome and time consuming. We were also able to perform constrained random verification very easily and systematically. From the study, we confirmed the various advantages of using the UVM register model in verification such as scalability, reusability and interoperability, and set some design guidelines for verification of the NPP controllers.

The End-To-End Safety Verification Process Implemented to Ensure Safe Operations of the Columbus Research Module

Science.gov (United States)

Arndt, J.; Kreimer, J.

2010-09-01

The European Space Laboratory COLUMBUS was launched in February 2008 with NASA Space Shuttle Atlantis. Since successful docking and activation this manned laboratory forms part of the International Space Station(ISS). Depending on the objectives of the Mission Increments the on-orbit configuration of the COLUMBUS Module varies with each increment. This paper describes the end-to-end verification which has been implemented to ensure safe operations under the condition of a changing on-orbit configuration. That verification process has to cover not only the configuration changes as foreseen by the Mission Increment planning but also those configuration changes on short notice which become necessary due to near real-time requests initiated by crew or Flight Control, and changes - most challenging since unpredictable - due to on-orbit anomalies. Subject of the safety verification is on one hand the on orbit configuration itself including the hardware and software products, on the other hand the related Ground facilities needed for commanding of and communication to the on-orbit System. But also the operational products, e.g. the procedures prepared for crew and ground control in accordance to increment planning, are subject of the overall safety verification. In order to analyse the on-orbit configuration for potential hazards and to verify the implementation of the related Safety required hazard controls, a hierarchical approach is applied. The key element of the analytical safety integration of the whole COLUMBUS Payload Complement including hardware owned by International Partners is the Integrated Experiment Hazard Assessment(IEHA). The IEHA especially identifies those hazardous scenarios which could potentially arise through physical and operational interaction of experiments. A major challenge is the implementation of a Safety process which owns quite some rigidity in order to provide reliable verification of on-board Safety and which likewise provides enough

Compositional Safety Analysis using Barrier Certificates

DEFF Research Database (Denmark)

Sloth, Christoffer; Pappas, George J.; Wisniewski, Rafael

2012-01-01

This paper proposes a compositional method for verifying the safety of a dynamical system, given as an interconnection of subsystems. The safety verification is conducted by the use of the barrier certificate method; hence, the contribution of this paper is to show how to obtain compositional...... conditions for safety verification. We show how to formulate the verification problem, as a composition of coupled subproblems, each given for one subsystem. Furthermore, we show how to find the compositional barrier certificates via linear and sum of squares programming problems. The proposed method makes...... it possible to verify the safety of higher dimensional systems, than the method for centrally computed barrier certificates. This is demonstrated by verifying the safety of an emergency shutdown of a wind turbine....

Design verification enhancement of field programmable gate array-based safety-critical I&C system of nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Ahmed, Ibrahim [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104 (Korea, Republic of); Jung, Jaecheon, E-mail: jcjung@kings.ac.kr [Department of Nuclear Power Plant Engineering, KEPCO International Nuclear Graduate School, 658-91 Haemaji-ro, Seosang-myeon, Ulju-gun, Ulsan 45014 (Korea, Republic of); Heo, Gyunyoung [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104 (Korea, Republic of)

2017-06-15

Highlights: • An enhanced, systematic and integrated design verification approach is proposed for V&V of FPGA-based I&C system of NPP. • RPS bistable fixed setpoint trip algorithm is designed, analyzed, verified and discussed using the proposed approaches. • The application of integrated verification approach simultaneously verified the entire design modules. • The applicability of the proposed V&V facilitated the design verification processes. - Abstract: Safety-critical instrumentation and control (I&C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. However, safety analysis for FPGA-based I&C systems, and verification and validation (V&V) assessments still remain important issues to be resolved, which are now become a global research point of interests. In this work, we proposed a systematic design and verification strategies from start to ready-to-use in form of model-based approaches for FPGA-based reactor protection system (RPS) that can lead to the enhancement of the design verification and validation processes. The proposed methodology stages are requirement analysis, enhanced functional flow block diagram (EFFBD) models, finite state machine with data path (FSMD) models, hardware description language (HDL) code development, and design verifications. The design verification stage includes unit test – Very high speed integrated circuit Hardware Description Language (VHDL) test and modified condition decision coverage (MC/DC) test, module test – MATLAB/Simulink Co-simulation test, and integration test – FPGA hardware test beds. To prove the adequacy of the proposed

Design verification enhancement of field programmable gate array-based safety-critical I&C system of nuclear power plant

International Nuclear Information System (INIS)

Ahmed, Ibrahim; Jung, Jaecheon; Heo, Gyunyoung

2017-01-01

Highlights: • An enhanced, systematic and integrated design verification approach is proposed for V&V of FPGA-based I&C system of NPP. • RPS bistable fixed setpoint trip algorithm is designed, analyzed, verified and discussed using the proposed approaches. • The application of integrated verification approach simultaneously verified the entire design modules. • The applicability of the proposed V&V facilitated the design verification processes. - Abstract: Safety-critical instrumentation and control (I&C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. However, safety analysis for FPGA-based I&C systems, and verification and validation (V&V) assessments still remain important issues to be resolved, which are now become a global research point of interests. In this work, we proposed a systematic design and verification strategies from start to ready-to-use in form of model-based approaches for FPGA-based reactor protection system (RPS) that can lead to the enhancement of the design verification and validation processes. The proposed methodology stages are requirement analysis, enhanced functional flow block diagram (EFFBD) models, finite state machine with data path (FSMD) models, hardware description language (HDL) code development, and design verifications. The design verification stage includes unit test – Very high speed integrated circuit Hardware Description Language (VHDL) test and modified condition decision coverage (MC/DC) test, module test – MATLAB/Simulink Co-simulation test, and integration test – FPGA hardware test beds. To prove the adequacy of the proposed

77 FR 50723 - Verification, Validation, Reviews, and Audits for Digital Computer Software Used in Safety...

Science.gov (United States)

2012-08-22

... regulations with respect to software verification and auditing of digital computer software used in the safety... Standards and Records,'' which requires, in part, that a quality assurance program be established and implemented to provide adequate assurance that systems and components important to safety will satisfactorily...

A Particle System for Safety Verification of Free Flight in Air Traffic

NARCIS (Netherlands)

Blom, H.A.P.; Krystul, J.; Bakker, G.J.

2006-01-01

Under free flight, an aircrew has both the freedom to select their trajectory and the responsibility of resolving conflicts with other aircraft. The general belief is that free flight can be made safe under low traffic conditions. Increasing traffic, however, raises safety verification issues. This

The dynamic flowgraph methodology as a safety analysis tool : programmable electronic system design and verification

NARCIS (Netherlands)

Houtermans, M.J.M.; Apostolakis, G.E.; Brombacher, A.C.; Karydas, D.M.

2002-01-01

The objective of this paper is to demonstrate the use of the Dynamic Flowgraph Methodology (DFM) during the design and verification of programmable electronic safety-related systems. The safety system consists of hardware as well as software. This paper explains and demonstrates the use of DFM, and

Verification and validation benchmarks

International Nuclear Information System (INIS)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-01-01

Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

Verification and validation benchmarks

International Nuclear Information System (INIS)

Oberkampf, William L.; Trucano, Timothy G.

2008-01-01

Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

Verification and Examination Management of Complex Systems

Directory of Open Access Journals (Sweden)

Stian Ruud

2014-10-01

Full Text Available As ship systems become more complex, with an increasing number of safety-critical functions, many interconnected subsystems, tight integration to other systems, and a large amount of potential failure modes, several industry parties have identified the need for improved methods for managing the verification and examination efforts of such complex systems. Such needs are even more prominent now that the marine and offshore industries are targeting more activities and operations in the Arctic environment. In this paper, a set of requirements and a method for verification and examination management are proposed for allocating examination efforts to selected subsystems. The method is based on a definition of a verification risk function for a given system topology and given requirements. The marginal verification risks for the subsystems may then be evaluated, so that examination efforts for the subsystem can be allocated. Two cases of requirements and systems are used to demonstrate the proposed method. The method establishes a systematic relationship between the verification loss, the logic system topology, verification method performance, examination stop criterion, the required examination effort, and a proposed sequence of examinations to reach the examination stop criterion.

Verification of BGA type FPGA logic applied to a control equipment with Safety Class using the special socket

International Nuclear Information System (INIS)

Chung, YounHu; Yoo, Kwanwoo; Lee, Myeongkyun; Yun, Donghwa

2015-01-01

This article aims to provide the verification method for BGA-type FPGA of Programmable Logic Controller (PLC) developed as Safety Class. The logic of FPGA in the control device with Safety Class is the circuit to control overall logic of PLC. This device converts to the different module from the input signals for both digital and analogue of the equipment in the field and outputs their data. In addition, it should perform the logical controls such as backplane communication control and data communication. We suggest acquiring method of the data signal with efficient logic using the socket in this article. Proposed test socket is made by simpler process than former one, and the process is done in batches by which cost can be reduces, and the test socket can be quickly produced in response to any request. Also, it is possible to reduce the wear by reducing the contact force of the ball phenomenon. The structure on the basis of silicon can be reduced the modification, and it has excellent linearity. At the logic verification, the operation that state data block is designed in the FPGA could be easily confirmed by using a socket

Verification and validation issues for digitally-based NPP safety systems

International Nuclear Information System (INIS)

Ets, A.R.

1993-01-01

The trend toward standardization, integration and reduced costs has led to increasing use of digital systems in reactor protection systems. While digital systems provide maintenance and performance advantages, their use also introduces new safety issues, in particular with regard to software. Current practice relies on verification and validation (V and V) to ensure the quality of safety software. However, effective V and V must be done in conjunction with a structured software development process and must consider the context of the safety system application. This paper present some of the issues and concerns that impact on the V and V process. These include documentation of systems requirements, common mode failures, hazards analysis and independence. These issues and concerns arose during evaluations of NPP safety systems for advanced reactor designs and digital I and C retrofits for existing nuclear plants in the United States. The pragmatic lessons from actual systems reviews can provide a basis for further refinement and development of guidelines for applying V and V to NPP safety systems. (author). 14 refs

Technique for unit testing of safety software verification and validation

International Nuclear Information System (INIS)

Li Duo; Zhang Liangju; Feng Junting

2008-01-01

The key issue arising from digitalization of the reactor protection system for nuclear power plant is how to carry out verification and validation (V and V), to demonstrate and confirm the software that performs reactor safety functions is safe and reliable. One of the most important processes for software V and V is unit testing, which verifies and validates the software coding based on concept design for consistency, correctness and completeness during software development. The paper shows a preliminary study on the technique for unit testing of safety software V and V, focusing on such aspects as how to confirm test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed here was successfully used in the work of unit testing on safety software of a digital reactor protection system. (authors)

DOE handbook: Integrated safety management systems (ISMS) verification. Team leader's handbook

International Nuclear Information System (INIS)

1999-06-01

The primary purpose of this handbook is to provide guidance to the ISMS verification Team Leader and the verification team in conducting ISMS verifications. The handbook describes methods and approaches for the review of the ISMS documentation (Phase I) and ISMS implementation (Phase II) and provides information useful to the Team Leader in preparing the review plan, selecting and training the team, coordinating the conduct of the verification, and documenting the results. The process and techniques described are based on the results of several pilot ISMS verifications that have been conducted across the DOE complex. A secondary purpose of this handbook is to provide information useful in developing DOE personnel to conduct these reviews. Specifically, this handbook describes methods and approaches to: (1) Develop the scope of the Phase 1 and Phase 2 review processes to be consistent with the history, hazards, and complexity of the site, facility, or activity; (2) Develop procedures for the conduct of the Phase 1 review, validating that the ISMS documentation satisfies the DEAR clause as amplified in DOE Policies 450.4, 450.5, 450.6 and associated guidance and that DOE can effectively execute responsibilities as described in the Functions, Responsibilities, and Authorities Manual (FRAM); (3) Develop procedures for the conduct of the Phase 2 review, validating that the description approved by the Approval Authority, following or concurrent with the Phase 1 review, has been implemented; and (4) Describe a methodology by which the DOE ISMS verification teams will be advised, trained, and/or mentored to conduct subsequent ISMS verifications. The handbook provides proven methods and approaches for verifying that commitments related to the DEAR, the FRAM, and associated amplifying guidance are in place and implemented in nuclear and high risk facilities. This handbook also contains useful guidance to line managers when preparing for a review of ISMS for radiological

Formal verification of complex properties on PLC programs

CERN Document Server

Darvas, D; Voros, A; Bartha, T; Blanco Vinuela, E; Gonzalez Suarez, V M

2014-01-01

Formal verification has become a recommended practice in the safety-critical application areas. However, due to the complexity of practical control and safety systems, the state space explosion often prevents the use of formal analysis. In this paper we extend our former verification methodology with effective property preserving reduction techniques. For this purpose we developed general rule-based reductions and a customized version of the Cone of Influence (COI) reduction. Using these methods, the verification of complex requirements formalised with temporal logics (e.g. CTL, LTL) can be orders of magnitude faster. We use the NuSMV model checker on a real-life PLC program from CERN to demonstrate the performance of our reduction techniques.

Improving Patient Safety With Error Identification in Chemotherapy Orders by Verification Nurses.

Science.gov (United States)

Baldwin, Abigail; Rodriguez, Elizabeth S

2016-02-01

The prevalence of medication errors associated with chemotherapy administration is not precisely known. Little evidence exists concerning the extent or nature of errors; however, some evidence demonstrates that errors are related to prescribing. This article demonstrates how the review of chemotherapy orders by a designated nurse known as a verification nurse (VN) at a National Cancer Institute-designated comprehensive cancer center helps to identify prescribing errors that may prevent chemotherapy administration mistakes and improve patient safety in outpatient infusion units. This article will describe the role of the VN and details of the verification process. To identify benefits of the VN role, a retrospective review and analysis of chemotherapy near-miss events from 2009-2014 was performed. A total of 4,282 events related to chemotherapy were entered into the Reporting to Improve Safety and Quality system. A majority of the events were categorized as near-miss events, or those that, because of chance, did not result in patient injury, and were identified at the point of prescribing.

Taiwan Power Company's power distribution analysis and fuel thermal margin verification methods for pressurized water reactors

International Nuclear Information System (INIS)

Huang, P.H.

1995-01-01

Taiwan Power Company's (TPC's) power distribution analysis and fuel thermal margin verification methods for pressurized water reactors (PWRs) are examined. The TPC and the Institute of Nuclear Energy Research started a joint 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, these methods were developed to allow TPC to independently perform verifications of the local power density and departure from nucleate boiling design bases, which are required by the reload safety evaluation for the Maanshan PWR plant. The computer codes utilized were extensively validated for the intended applications. Sample calculations were performed for up to six reload cycles of the Maanshan plant, and the results were found to be quite consistent with the vendor's calculational results

Training to Support Standardization and Improvement of Safety I and C Related Verification and Validation Activities

Energy Technology Data Exchange (ETDEWEB)

Ammon, G.; Schoenfelder, C.

2014-07-01

In recent years AREVA has conducted several measures to enhance the effectiveness of safety I and C related verification and validation activities within nuclear power plant (NPP) new build as well as modernization projects, thereby further strengthening its commitment to achieving the highest level of safety in nuclear facilities. (Author)

Verification of criticality safety in on-site spent fuel storage systems

International Nuclear Information System (INIS)

Rasmussen, R.W.

1989-01-01

On February 15, 1984, Duke Power Company received approval for a two-region, burnup credit, spent fuel storage rack design at both Units 1 and 2 of the McGuire Nuclear Station. Duke also hopes to obtain approval by January of 1990 for a dry spent fuel storage system at the Oconee Nuclear Station, which will incorporate the use of burnup credit in the criticality analysis governing the design of the individual storage units. While experiences in burnup verification for criticality safety for their dry storage system at Oconee are in the future, the methods proposed for burnup verification will be similar to those currently used at the McGuire Nuclear Station in the two-region storage racks installed in both pools. In conclusion, the primary benefit of the McGuire rerack effort has obviously been the amount of storage expansion it provided. A total increase of about 2,000 storage cells was realized, 1,000 of which were the result of pursuing the two-region rather than the conventional poison rack design. Less impacting, but equally as important, however, has been the experience gained during the planning, installation, and operation of these storage racks. This experience should prove useful for future rerack efforts likely to occur at Duke's Catawba Nuclear Station as well as for the current dry storage effort underway for the Oconee Nuclear Station

Analysis using formal method and testing technique for the processor module for safety-critical application

International Nuclear Information System (INIS)

Choi, J. Y.; Choi, B. J.; Song, H. J.; Hwang, D. Y.; Song, G. H.; Lee, H.

2008-06-01

This research is on help develop nuclear power plant control system, through the requirement specification and verification method development. As the result of applying the test method, a test standard was obtain through test documentation writing support and a test document reflecting the standard test activities based on the test standard. The specification and verification of the pCOS system and the unified testing documentation and execution helps the entire project to progress and enable us to achieve necessary documents and technology to develop a safety critical system

Analysis using formal method and testing technique for the processor module for safety-critical application

Energy Technology Data Exchange (ETDEWEB)

Choi, J. Y.; Choi, B. J.; Song, H. J.; Hwang, D. Y.; Song, G. H.; Lee, H. [Korea University, Seoul (Korea, Republic of)

2008-06-15

This research is on help develop nuclear power plant control system, through the requirement specification and verification method development. As the result of applying the test method, a test standard was obtain through test documentation writing support and a test document reflecting the standard test activities based on the test standard. The specification and verification of the pCOS system and the unified testing documentation and execution helps the entire project to progress and enable us to achieve necessary documents and technology to develop a safety critical system.

Electric and mechanical basic parameters to elaborate a process for a technical verification of safety related design modifications

International Nuclear Information System (INIS)

Lamuno Fernandez, Mercedes; La Roca Mallofre, GISEL; Bano Azcon, Alberto

2010-01-01

This paper presents a systematic process to check a design in order to achieve all the requirements that regulations demand. Nuclear engineers must verify that a design is done according to the safety requirements, and this paper presents how we have elaborated a process to improve the technical project verification. For a faster, better and easier verification process, here we summarize how to select the electric and mechanical basic parameters, which ensure the correct project verification of safety related design modifications. This process considers different aspects, which guarantee that the design preserves the availability, reliability and functional capability of the Structures, Systems and Components needed to operate the Nuclear Power Station with security. Electric and mechanical reference parameters are identified and discussed as well as others related ones, which are critical to safety. The implementation procedure to develop tasks performed in any company that has a quality plan is a requirement. On the engineering business, it is important not to use the personal criteria to do a technical analysis of a project; although, many times it is the checker's criteria and knowledge responsibility to ensure the correct development of a design modification. Then, the checker capabilities are the basis of the modification verification. This kind of procedure's development is not easy, because in an engineering project with important technical contents, there are multiple scenarios, but lots of them have a common basis. If we can identify the technical common basis of these projects, we will make good project verification but there are many difficulties we can encounter along this process. (authors)

Neutron spectrometric methods for core inventory verification in research reactors

International Nuclear Information System (INIS)

Ellinger, A.; Filges, U.; Hansen, W.; Knorr, J.; Schneider, R.

2002-01-01

In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors

Research on neutron source multiplication method in nuclear critical safety

International Nuclear Information System (INIS)

Zhu Qingfu; Shi Yongqian; Hu Dingsheng

2005-01-01

The paper concerns in the neutron source multiplication method research in nuclear critical safety. Based on the neutron diffusion equation with external neutron source the effective sub-critical multiplication factor k s is deduced, and k s is different to the effective neutron multiplication factor k eff in the case of sub-critical system with external neutron source. The verification experiment on the sub-critical system indicates that the parameter measured with neutron source multiplication method is k s , and k s is related to the external neutron source position in sub-critical system and external neutron source spectrum. The relation between k s and k eff and the effect of them on nuclear critical safety is discussed. (author)

Java bytecode verification via static single assignment form

DEFF Research Database (Denmark)

Gal, Andreas; Probst, Christian W.; Franz, Michael

2008-01-01

Java Virtual Machines (JVMs) traditionally perform bytecode verification by way of an iterative data-flow analysis. Bytecode verification is necessary to ensure type safety because temporary variables in the JVM are not statically typed. We present an alternative verification mechanism that trans......Java Virtual Machines (JVMs) traditionally perform bytecode verification by way of an iterative data-flow analysis. Bytecode verification is necessary to ensure type safety because temporary variables in the JVM are not statically typed. We present an alternative verification mechanism...

International exchange on nuclear safety related expert systems: The role of software verification and validation

International Nuclear Information System (INIS)

Sun, B.K.H.

1996-01-01

An important lesson learned from the Three Mile Island accident is that human errors can be significant contributors to risk. Recent advancement in computer hardware and software technology helped make expert system techniques potentially viable tools for improving nuclear power plant safety and reliability. As part of the general man-machine interface technology, expert systems have recently become increasingly prominent as a potential solution to a number of previously intractable problems in many phases of human activity, including operation, maintenance, and engineering functions. Traditional methods for testing and analyzing analog systems are no longer adequate to handle the increased complexity of software systems. The role of Verification and Validation (V and V) is to add rigor to the software development and maintenance cycle to guarantee the high level confidence needed for applications. Verification includes the process and techniques for confirming that all the software requirements in one stage of the development are met before proceeding on to the next stage. Validation involves testing the integrated software and hardware system to ensure that it reliably fulfills its intended functions. Only through a comprehensive V and V program can a high level of confidence be achieved. There exist many different standards and techniques for software verification and validation, yet they lack uniform approaches that provides adequate levels of practical guidance which can be used by users for nuclear power plant applications. There is a need to unify different approaches for addressing software verification and validation and to develop practical and cost effective guidelines for user and regulatory acceptance. (author). 8 refs

Verification for excess reactivity on beginning equilibrium core of RSG GAS

International Nuclear Information System (INIS)

Daddy Setyawan; Budi Rohman

2011-01-01

BAPETEN is an institution authorized to control the use of nuclear energy in Indonesia. Control for the use of nuclear energy is carried out through three pillars: regulation, licensing, and inspection. In order to assure the safety of the operating research reactors, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the Power Peaking Factor in the equilibrium silicide core of RSG GAS reactor by computational method using MCNP-ORIGEN. This verification calculation results for is 9.4 %. Meanwhile, the RSG-GAS safety analysis report shows that the excess reactivity on equilibrium core of RSG GAS is 9.7 %. The verification calculation results show a good agreement with the report. (author)

Finite Countermodel Based Verification for Program Transformation (A Case Study

Directory of Open Access Journals (Sweden)

Alexei P. Lisitsa

2015-12-01

Full Text Available Both automatic program verification and program transformation are based on program analysis. In the past decade a number of approaches using various automatic general-purpose program transformation techniques (partial deduction, specialization, supercompilation for verification of unreachability properties of computing systems were introduced and demonstrated. On the other hand, the semantics based unfold-fold program transformation methods pose themselves diverse kinds of reachability tasks and try to solve them, aiming at improving the semantics tree of the program being transformed. That means some general-purpose verification methods may be used for strengthening program transformation techniques. This paper considers the question how finite countermodels for safety verification method might be used in Turchin's supercompilation method. We extract a number of supercompilation sub-algorithms trying to solve reachability problems and demonstrate use of an external countermodel finder for solving some of the problems.

Anchorage of equipment - requirements and verification methods with emphasis on equipment of existing and constructed VVER-type nuclear power plants

International Nuclear Information System (INIS)

Masopust, R.

1999-01-01

Criteria and verification methods which are recommended for use in the capacity evaluation of anchorage of safety-related equipment at WWER-type nuclear power plants are presented. Developed in compliance with the relevant basic standards documents specifically for anchorage of WWER-type equipment components, the criteria and methods cover different types of anchor bolts and other anchorage elements which are typical of existing, constructed, or reconstructed WWER-type nuclear power plants

CESAR cost-efficient methods and processes for safety-relevant embedded systems

CERN Document Server

Wahl, Thomas

2013-01-01

The book summarizes the findings and contributions of the European ARTEMIS project, CESAR, for improving and enabling interoperability of methods, tools, and processes to meet the demands in embedded systems development across four domains - avionics, automotive, automation, and rail. The contributions give insight to an improved engineering and safety process life-cycle for the development of safety critical systems. They present new concept of engineering tools integration platform to improve the development of safety critical embedded systems and illustrate capacity of this framework for end-user instantiation to specific domain needs and processes. They also advance state-of-the-art in component-based development as well as component and system validation and verification, with tool support. And finally they describe industry relevant evaluated processes and methods especially designed for the embedded systems sector as well as easy adoptable common interoperability principles for software tool integratio...

Formal development and verification of a distributed railway control system

DEFF Research Database (Denmark)

Haxthausen, Anne Elisabeth; Peleska, J.

2000-01-01

specifications which are transformed into directly implementable distributed control processes by applying a series of refinement and verification steps. Concrete safety requirements are derived from an abstract version that can be easily validated with respect to soundness and completeness. Complexity......The authors introduce the concept for a distributed railway control system and present the specification and verification of the main algorithm used for safe distributed control. Our design and verification approach is based on the RAISE method, starting with highly abstract algebraic...... is further reduced by separating the system model into a domain model and a controller model. The domain model describes the physical system in absence of control and the controller model introduces the safety-related control mechanisms as a separate entity monitoring observables of the physical system...

The inverse method parametric verification of real-time embedded systems

CERN Document Server

André , Etienne

2013-01-01

This book introduces state-of-the-art verification techniques for real-time embedded systems, based on the inverse method for parametric timed automata. It reviews popular formalisms for the specification and verification of timed concurrent systems and, in particular, timed automata as well as several extensions such as timed automata equipped with stopwatches, linear hybrid automata and affine hybrid automata.The inverse method is introduced, and its benefits for guaranteeing robustness in real-time systems are shown. Then, it is shown how an iteration of the inverse method can solv

Probabilistic Requirements (Partial) Verification Methods Best Practices Improvement. Variables Acceptance Sampling Calculators: Empirical Testing. Volume 2

Science.gov (United States)

Johnson, Kenneth L.; White, K. Preston, Jr.

2012-01-01

The NASA Engineering and Safety Center was requested to improve on the Best Practices document produced for the NESC assessment, Verification of Probabilistic Requirements for the Constellation Program, by giving a recommended procedure for using acceptance sampling by variables techniques as an alternative to the potentially resource-intensive acceptance sampling by attributes method given in the document. In this paper, the results of empirical tests intended to assess the accuracy of acceptance sampling plan calculators implemented for six variable distributions are presented.

Application of Integrated Verification Approach to FPGA-based Safety-Critical I and C System of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Ahmed, Ibrahim; Heo, Gyunyoung [Kyunghee Univ., Yongin (Korea, Republic of); Jung, Jaecheon [KEPCO, Ulsan (Korea, Republic of)

2016-10-15

Safety-critical instrumentation and control (I and C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. Generally in FPGA design verification, the designers make use of verification techniques by writing the test benches which involved various stages of verification activities of register-transfer level (RTL), gate-level, and place and route. Writing the test benches is considerably time consuming and require a lot of efforts to achieve a satisfied desire results. Furthermore, performing the verification at each stage is a major bottleneck and demanded much activities and time. In addition, verification is conceivably, the most difficult and complicated aspect of any design. Therefore, in view of these, this work applied an integrated verification approach to the verification of FPGA-based I and C system in NPP that simultaneously verified the whole design modules using MATLAB/Simulink HDL Co-simulation models. Verification is conceivably, the most difficult and complicated aspect of any design, and an FPGA design is not an exception. Therefore, in this work, we introduced and discussed how an application of integrated verification technique to the verification and testing of FPGA-based I and C system design in NPP can facilitate the verification processes, and verify the entire design modules of the system simultaneously using MATLAB/Simulink HDL co-simulation models. In conclusion, the results showed that, the integrated verification approach through MATLAB/Simulink models, if applied to any design to be verified, could speed up the design verification and reduce the V and V tasks.

Application of Integrated Verification Approach to FPGA-based Safety-Critical I and C System of Nuclear Power Plant

International Nuclear Information System (INIS)

Ahmed, Ibrahim; Heo, Gyunyoung; Jung, Jaecheon

2016-01-01

Safety-critical instrumentation and control (I and C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. Generally in FPGA design verification, the designers make use of verification techniques by writing the test benches which involved various stages of verification activities of register-transfer level (RTL), gate-level, and place and route. Writing the test benches is considerably time consuming and require a lot of efforts to achieve a satisfied desire results. Furthermore, performing the verification at each stage is a major bottleneck and demanded much activities and time. In addition, verification is conceivably, the most difficult and complicated aspect of any design. Therefore, in view of these, this work applied an integrated verification approach to the verification of FPGA-based I and C system in NPP that simultaneously verified the whole design modules using MATLAB/Simulink HDL Co-simulation models. Verification is conceivably, the most difficult and complicated aspect of any design, and an FPGA design is not an exception. Therefore, in this work, we introduced and discussed how an application of integrated verification technique to the verification and testing of FPGA-based I and C system design in NPP can facilitate the verification processes, and verify the entire design modules of the system simultaneously using MATLAB/Simulink HDL co-simulation models. In conclusion, the results showed that, the integrated verification approach through MATLAB/Simulink models, if applied to any design to be verified, could speed up the design verification and reduce the V and V tasks

Wu’s Characteristic Set Method for SystemVerilog Assertions Verification

Directory of Open Access Journals (Sweden)

Xinyan Gao

2013-01-01

Full Text Available We propose a verification solution based on characteristic set of Wu’s method towards SystemVerilog assertion checking over digital circuit systems. We define a suitable subset of SVAs so that an efficient polynomial modeling mechanism for both circuit descriptions and assertions can be applied. We present an algorithm framework based on the algebraic representations using characteristic set of polynomial system. This symbolic algebraic approach is a useful supplement to the existent verification methods based on simulation.

Algebraic Verification Method for SEREs Properties via Groebner Bases Approaches

Directory of Open Access Journals (Sweden)

Ning Zhou

2013-01-01

Full Text Available This work presents an efficient solution using computer algebra system to perform linear temporal properties verification for synchronous digital systems. The method is essentially based on both Groebner bases approaches and symbolic simulation. A mechanism for constructing canonical polynomial set based symbolic representations for both circuit descriptions and assertions is studied. We then present a complete checking algorithm framework based on these algebraic representations by using Groebner bases. The computational experience result in this work shows that the algebraic approach is a quite competitive checking method and will be a useful supplement to the existent verification methods based on simulation.

Design verification methodology for a solenoid valve for industrial applications

International Nuclear Information System (INIS)

Park, Chang Dae; Lim, Byung Ju; Chun, Kyung Yul

2015-01-01

Solenoid operated valves (SOV) are widely used in many applications due to their fast dynamic responses, cost effectiveness, and less contamination sensitive characteristics. In this paper, we tried to provide a convenient method of design verification of SOV to design engineers who depend on their experiences and experiment during design and development process of SOV. First, we summarize a detailed procedure for designing SOVs for industrial applications. All of the design constraints are defined in the first step of the design, and then the detail design procedure is presented based on design experiences as well as various physical and electromagnetic relationships. Secondly, we have suggested a verification method of this design using theoretical relationships, which enables optimal design of SOV from a point of view of safety factor of design attraction force. Lastly, experimental performance tests using several prototypes manufactured based on this design method show that the suggested design verification methodology is appropriate for designing new models of solenoids. We believe that this verification process is novel logic and useful to save time and expenses during development of SOV because verification tests with manufactured specimen may be substituted partly by this verification methodology.

The design and verification of probabilistic safety analysis platform NFRisk

International Nuclear Information System (INIS)

Hu Wenjun; Song Wei; Ren Lixia; Qian Hongtao

2010-01-01

To increase the technical ability in Probabilistic Safety Analysis (PSA) field in China,it is necessary and important to study and develop indigenous professional PSA platform. Following such principle as 'from structure simplification to modulization to production of cut sets to minimum of cut sets', the algorithms, including simplification algorithm, modulization algorithm, the algorithm of conversion from fault tree to binary decision diagram (BDD), the solving algorithm of cut sets, the minimum algorithm of cut sets, and so on, were designed and developed independently; the design of data management and operation platform was completed all alone; the verification and validation of NFRisk platform based on 3 typical fault trees was finished on our own. (authors)

IDEF method for designing seismic information system in CTBT verification

International Nuclear Information System (INIS)

Zheng Xuefeng; Shen Junyi; Jin Ping; Zhang Huimin; Zheng Jiangling; Sun Peng

2004-01-01

Seismic information system is of great importance for improving the capability of CTBT verification. A large amount of money has been appropriated for the research in this field in the U.S. and some other countries in recent years. However, designing and developing a seismic information system involves various technologies about complex system design. This paper discusses the IDEF0 method to construct function models and the IDEF1x method to make information models systemically, as well as how they are used in designing seismic information system in CTBT verification. (authors)

Embedded software verification and debugging

CERN Document Server

Winterholer, Markus

2017-01-01

This book provides comprehensive coverage of verification and debugging techniques for embedded software, which is frequently used in safety critical applications (e.g., automotive), where failures are unacceptable. Since the verification of complex systems needs to encompass the verification of both hardware and embedded software modules, this book focuses on verification and debugging approaches for embedded software with hardware dependencies. Coverage includes the entire flow of design, verification and debugging of embedded software and all key approaches to debugging, dynamic, static, and hybrid verification. This book discusses the current, industrial embedded software verification flow, as well as emerging trends with focus on formal and hybrid verification and debugging approaches. Includes in a single source the entire flow of design, verification and debugging of embedded software; Addresses the main techniques that are currently being used in the industry for assuring the quality of embedded softw...

Development and Verification of the Computer Codes for the Fast Reactors Nuclear Safety Justification

International Nuclear Information System (INIS)

Kisselev, A.E.; Mosunova, N.A.; Strizhov, V.F.

2015-01-01

The information on the status of the work on development of the system of the nuclear safety codes for fast liquid metal reactors is presented in paper. The purpose of the work is to create an instrument for NPP neutronic, thermohydraulic and strength justification including human and environment radiation safety. The main task that is to be solved by the system of codes developed is the analysis of the broad spectrum of phenomena taking place on the NPP (including reactor itself, NPP components, containment rooms, industrial site and surrounding area) and analysis of the impact of the regular and accidental releases on the environment. The code system is oriented on the ability of fully integrated modeling of the NPP behavior in the coupled definition accounting for the wide range of significant phenomena taking place on the NPP under normal and accident conditions. It is based on the models that meet the state-of-the-art knowledge level. The codes incorporate advanced numerical methods and modern programming technologies oriented on the high-performance computing systems. The information on the status of the work on verification of the separate codes of the system of codes is also presented. (author)

Automatic Verification of Timing Constraints for Safety Critical Space Systems

Science.gov (United States)

Fernandez, Javier; Parra, Pablo; Sanchez Prieto, Sebastian; Polo, Oscar; Bernat, Guillem

2015-09-01

In this paper is presented an automatic process of verification. We focus in the verification of scheduling analysis parameter. This proposal is part of process based on Model Driven Engineering to automate a Verification and Validation process of the software on board of satellites. This process is implemented in a software control unit of the energy particle detector which is payload of Solar Orbiter mission. From the design model is generated a scheduling analysis model and its verification model. The verification as defined as constraints in way of Finite Timed Automatas. When the system is deployed on target the verification evidence is extracted as instrumented points. The constraints are fed with the evidence, if any of the constraints is not satisfied for the on target evidence the scheduling analysis is not valid.

Application of verification and validation on safety parameter display systems

International Nuclear Information System (INIS)

Thomas, N.C.

1983-01-01

Offers some explanation of how verification and validation (VandV) can support development and licensing of the Safety Parameter Display Systems (SPDS). Advocates that VandV can be more readily accepted within the nuclear industry if a better understanding exists of what the objectives of VandV are and should be. Includes a discussion regarding a reasonable balance of costs and benefits of VandV as applied to the SPDS and to other digital systems. Represents the author's perception of the regulator's perspective based on background information and experience, and discussions with regulators about their current concerns and objectives. Suggests that the introduction of the SPDS into the Control Room is a first step towards growing dependency on use of computers

The MODUS Approach to Formal Verification

Directory of Open Access Journals (Sweden)

Brewka Lukasz

2014-03-01

Full Text Available Background: Software reliability is of great importance for the development of embedded systems that are often used in applications that have requirements for safety. Since the life cycle of embedded products is becoming shorter, productivity and quality simultaneously required and closely in the process of providing competitive products Objectives: In relation to this, MODUS (Method and supporting toolset advancing embedded systems quality project aims to provide small and medium-sized businesses ways to improve their position in the embedded market through a pragmatic and viable solution Methods/Approach: This paper will describe the MODUS project with focus on the technical methodologies that can assist formal verification and formal model checking. Results: Based on automated analysis of the characteristics of the system and by controlling the choice of the existing opensource model verification engines, model verification producing inputs to be fed into these engines. Conclusions: The MODUS approach is aligned with present market needs; the familiarity with tools, the ease of use and compatibility/interoperability remain among the most important criteria when selecting the development environment for a project

Multilateral disarmament verification

International Nuclear Information System (INIS)

Persbo, A.

2013-01-01

Non-governmental organisations, such as VERTIC (Verification Research, Training and Information Centre), can play an important role in the promotion of multilateral verification. Parties involved in negotiating nuclear arms accords are for the most part keen that such agreements include suitable and robust provisions for monitoring and verification. Generally progress in multilateral arms control verification is often painstakingly slow, but from time to time 'windows of opportunity' - that is, moments where ideas, technical feasibility and political interests are aligned at both domestic and international levels - may occur and we have to be ready, so the preparatory work is very important. In the context of nuclear disarmament, verification (whether bilateral or multilateral) entails an array of challenges, hurdles and potential pitfalls relating to national security, health, safety and even non-proliferation, so preparatory work is complex and time-greedy. A UK-Norway Initiative was established in order to investigate the role that a non-nuclear-weapon state such as Norway could potentially play in the field of nuclear arms control verification. (A.C.)

Thermal reactor safety

International Nuclear Information System (INIS)

1980-06-01

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport

Thermal reactor safety

Energy Technology Data Exchange (ETDEWEB)

1980-06-01

Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

International Nuclear Information System (INIS)

Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

2005-01-01

This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

Internal safety review team at Comanche Peak SES

Energy Technology Data Exchange (ETDEWEB)

Davis, D [Comanche Peak Steam Electric Staion, Texas Utilities, TX (United States)

1997-09-01

The presentations describes the following issues: levels of defense in depth; internal safety review organizations; methods used to perform safety assessment; safety committee review; quality verification; root cause analysis; human performance program; industry operating experience.

Verification and validation of the safety parameter display system for nuclear power plant

International Nuclear Information System (INIS)

Zhang Yuanfang

1993-05-01

During the design and development phase of the safety parameter display system for nuclear power plant, a verification and validation (V and V) plan has been implemented to improve the quality of system design. The V and V activities are briefly introduced, which were executed in four stages of feasibility research, system design, code development and system integration and regulation. The evaluation plan and the process of implementation as well as the evaluation conclusion of the final technical validation for this system are also presented in detail

Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

International Nuclear Information System (INIS)

Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

1994-01-01

Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

Software Safety and Security

CERN Document Server

Nipkow, T; Hauptmann, B

2012-01-01

Recent decades have seen major advances in methods and tools for checking the safety and security of software systems. Automatic tools can now detect security flaws not only in programs of the order of a million lines of code, but also in high-level protocol descriptions. There has also been something of a breakthrough in the area of operating system verification. This book presents the lectures from the NATO Advanced Study Institute on Tools for Analysis and Verification of Software Safety and Security; a summer school held at Bayrischzell, Germany, in 2011. This Advanced Study Institute was

Performance Verification for Safety Injection Tank with Fluidic Device

International Nuclear Information System (INIS)

Yune, Seok Jeong; Kim, Da Yong

2014-01-01

In LBLOCA, the SITs of a conventional nuclear power plant deliver excessive cooling water to the reactor vessel causing the water to flow into the containment atmosphere. In an effort to make it more efficient, Fluidic Device (FD) is installed inside a SIT of Advanced Power Reactor 1400 (APR 1400). FD, a complete passive controller which doesn't require actuating power, controls injection flow rates which are susceptible to a change in the flow resistance inside a vortex chamber of FD. When SIT Emergency Core Cooling (ECC) water level is above the top of the stand pipe, the water enters the vortex chamber through both the top of the stand pipe and the control ports resulting in injection of the water at a large flow rate. When the water level drops below the top of the stand pipe, the water only enters the vortex chamber through the control ports resulting in vortex formation in the vortex chamber and a relatively small flow injection. Performance verification of SIT shall be carried out because SITs play an integral role to mitigate accidents. In this paper, the performance verification method of SIT with FD is presented. In this paper, the equations for calculation of flow resistance coefficient (K) are induced to evaluate on-site performance of APR 1400 SIT with FD. Then, the equations are applied to the performance verification of SIT with FD and good results are obtained

WE-D-BRA-04: Online 3D EPID-Based Dose Verification for Optimum Patient Safety

International Nuclear Information System (INIS)

Spreeuw, H; Rozendaal, R; Olaciregui-Ruiz, I; Mans, A; Mijnheer, B; Herk, M van; Gonzalez, P

2015-01-01

Purpose: To develop an online 3D dose verification tool based on EPID transit dosimetry to ensure optimum patient safety in radiotherapy treatments. Methods: A new software package was developed which processes EPID portal images online using a back-projection algorithm for the 3D dose reconstruction. The package processes portal images faster than the acquisition rate of the portal imager (∼ 2.5 fps). After a portal image is acquired, the software seeks for “hot spots” in the reconstructed 3D dose distribution. A hot spot is in this study defined as a 4 cm 3 cube where the average cumulative reconstructed dose exceeds the average total planned dose by at least 20% and 50 cGy. If a hot spot is detected, an alert is generated resulting in a linac halt. The software has been tested by irradiating an Alderson phantom after introducing various types of serious delivery errors. Results: In our first experiment the Alderson phantom was irradiated with two arcs from a 6 MV VMAT H&N treatment having a large leaf position error or a large monitor unit error. For both arcs and both errors the linac was halted before dose delivery was completed. When no error was introduced, the linac was not halted. The complete processing of a single portal frame, including hot spot detection, takes about 220 ms on a dual hexacore Intel Xeon 25 X5650 CPU at 2.66 GHz. Conclusion: A prototype online 3D dose verification tool using portal imaging has been developed and successfully tested for various kinds of gross delivery errors. The detection of hot spots was proven to be effective for the timely detection of these errors. Current work is focused on hot spot detection criteria for various treatment sites and the introduction of a clinical pilot program with online verification of hypo-fractionated (lung) treatments

Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

International Nuclear Information System (INIS)

G. L. Sharp; R. T. McCracken

2004-01-01

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

Formal Development and Verification of Railway Control Systems - In the context of ERTMS/ETCS Level 2

DEFF Research Database (Denmark)

Vu, Linh Hong

This dissertation presents a holistic, formal method for efficient modelling and verification of safety-critical railway control systems that have product line characteristics, i.e., each individual system is constructed by instantiating common generic applications with concrete configuration dat...... standardized railway control systems ERTMS/ETCS Level 2. Experiments showed that the method can be used for specification, verification and validation of systems of industrial size....

Evaluation of verification methods for input-accountability measurements

International Nuclear Information System (INIS)

Maeck, W.J.

1980-01-01

As part of TASTEX related programs two independent methods have been evaluated for the purpose of providing verification of the amount of Pu charged to the head-end of a nuclear fuel processing plant. The first is the Pu/U (gravimetric method), TASTEX Task-L, and the second is the Tracer Method, designated Task-M. Summaries of the basic technology, results of various studies under actual plant conditions, future requirements, are given for each of the Tasks

A Feature Subtraction Method for Image Based Kinship Verification under Uncontrolled Environments

DEFF Research Database (Denmark)

Duan, Xiaodong; Tan, Zheng-Hua

2015-01-01

The most fundamental problem of local feature based kinship verification methods is that a local feature can capture the variations of environmental conditions and the differences between two persons having a kin relation, which can significantly decrease the performance. To address this problem...... the feature distance between face image pairs with kinship and maximize the distance between non-kinship pairs. Based on the subtracted feature, the verification is realized through a simple Gaussian based distance comparison method. Experiments on two public databases show that the feature subtraction method...

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

Energy Technology Data Exchange (ETDEWEB)

Matloch, L.; Vaccaro, S.; Couland, M.; De Baere, P.; Schwalbach, P. [Euratom, Communaute europeenne de l' energie atomique - CEEA (European Commission (EC))

2015-07-01

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

Validation, verification and evaluation of a Train to Train Distance Measurement System by means of Colored Petri Nets

International Nuclear Information System (INIS)

Song, Haifeng; Liu, Jieyu; Schnieder, Eckehard

2017-01-01

Validation, verification and evaluation are necessary processes to assure the safety and functionality of a system before its application in practice. This paper presents a Train to Train Distance Measurement System (TTDMS), which can provide distance information independently from existing onboard equipment. Afterwards, we proposed a new process using Colored Petri Nets to verify the TTDMS system functional safety, as well as to evaluate the system performance. Three main contributions are carried out in the paper: Firstly, this paper proposes a formalized TTDMS model, and the model correctness is validated using state space analysis and simulation-based verification. Secondly, corresponding checking queries are proposed for the purpose of functional safety verification. Further, the TTDMS performance is evaluated by applying parameters in the formal model. Thirdly, the reliability of a functional prototype TTDMS is estimated. It is found that the procedure can cooperate with the system development, and both formal and simulation-based verifications are performed. Using our process to evaluate and verify a system is easier to read and more reliable compared to executable code and mathematical methods. - Highlights: • A new Train to Train Distance Measurement System. • New approach verifying system functional safety and evaluating system performance by means of CPN. • System formalization using the system property concept. • Verification of system functional safety using state space analysis. • Evaluation of system performance applying simulation-based analysis.

Survey of Existing Tools for Formal Verification.

Energy Technology Data Exchange (ETDEWEB)

Punnoose, Ratish J.; Armstrong, Robert C.; Wong, Matthew H.; Jackson, Mayo

2014-12-01

Formal methods have come into wide use because of their effectiveness in verifying "safety and security" requirements of digital systems; a set of requirements for which testing is mostly ineffective. Formal methods are routinely used in the design and verification of high-consequence digital systems in industry. This report outlines our work in assessing the capabilities of commercial and open source formal tools and the ways in which they can be leveraged in digital design workflows.

Environmental technology verification methods

CSIR Research Space (South Africa)

Szewczuk, S

2016-03-01

Full Text Available Environmental Technology Verification (ETV) is a tool that has been developed in the United States of America, Europe and many other countries around the world to help innovative environmental technologies reach the market. Claims about...

Integrated Safety Management System Phase 1 and 2 Verification for the Environmental Restoration Contractor Volumes 1 and 2

Energy Technology Data Exchange (ETDEWEB)

CARTER, R.P.

2000-04-04

DOE Policy 450.4 mandates that safety be integrated into all aspects of the management and operations of its facilities. The goal of an institutionalized Integrated Safety Management System (ISMS) is to have a single integrated system that includes Environment, Safety, and Health requirements in the work planning and execution processes to ensure the protection of the worker, public, environment, and the federal property over the life cycle of the Environmental Restoration (ER) Project. The purpose of this Environmental Restoration Contractor (ERC) ISMS Phase MI Verification was to determine whether ISMS programs and processes were institutionalized within the ER Project, whether these programs and processes were implemented, and whether the system had promoted the development of a safety conscious work culture.

A study of compositional verification based IMA integration method

Science.gov (United States)

Huang, Hui; Zhang, Guoquan; Xu, Wanmeng

2018-03-01

The rapid development of avionics systems is driving the application of integrated modular avionics (IMA) systems. But meanwhile it is improving avionics system integration, complexity of system test. Then we need simplify the method of IMA system test. The IMA system supports a module platform that runs multiple applications, and shares processing resources. Compared with federated avionics system, IMA system is difficult to isolate failure. Therefore, IMA system verification will face the critical problem is how to test shared resources of multiple application. For a simple avionics system, traditional test methods are easily realizing to test a whole system. But for a complex system, it is hard completed to totally test a huge and integrated avionics system. Then this paper provides using compositional-verification theory in IMA system test, so that reducing processes of test and improving efficiency, consequently economizing costs of IMA system integration.

Model-based verification method for solving the parameter uncertainty in the train control system

International Nuclear Information System (INIS)

Cheng, Ruijun; Zhou, Jin; Chen, Dewang; Song, Yongduan

2016-01-01

This paper presents a parameter analysis method to solve the parameter uncertainty problem for hybrid system and explore the correlation of key parameters for distributed control system. For improving the reusability of control model, the proposed approach provides the support for obtaining the constraint sets of all uncertain parameters in the abstract linear hybrid automata (LHA) model when satisfying the safety requirements of the train control system. Then, in order to solve the state space explosion problem, the online verification method is proposed to monitor the operating status of high-speed trains online because of the real-time property of the train control system. Furthermore, we construct the LHA formal models of train tracking model and movement authority (MA) generation process as cases to illustrate the effectiveness and efficiency of the proposed method. In the first case, we obtain the constraint sets of uncertain parameters to avoid collision between trains. In the second case, the correlation of position report cycle and MA generation cycle is analyzed under both the normal and the abnormal condition influenced by packet-loss factor. Finally, considering stochastic characterization of time distributions and real-time feature of moving block control system, the transient probabilities of wireless communication process are obtained by stochastic time petri nets. - Highlights: • We solve the parameters uncertainty problem by using model-based method. • We acquire the parameter constraint sets by verifying linear hybrid automata models. • Online verification algorithms are designed to monitor the high-speed trains. • We analyze the correlation of key parameters and uncritical parameters. • The transient probabilities are obtained by using reliability analysis.

Numerical Verification Methods for Spherical $t$-Designs

OpenAIRE

Chen, Xiaojun

2009-01-01

The construction of spherical $t$-designs with $(t+1)^2$ points on the unit sphere $S^2$ in $\\mathbb{R}^3$ can be reformulated as an underdetermined system of nonlinear equations. This system is highly nonlinear and involves the evaluation of a degree $t$ polynomial in $(t+1)^4$ arguments. This paper reviews numerical verification methods using the Brouwer fixed point theorem and Krawczyk interval operator for solutions of the underdetermined system of nonlinear equations...

CFD code verification and the method of manufactured solutions

International Nuclear Information System (INIS)

Pelletier, D.; Roache, P.J.

2002-01-01

This paper presents the Method of Manufactured Solutions (MMS) for CFD code verification. The MMS provides benchmark solutions for direct evaluation of the solution error. The best benchmarks are exact analytical solutions with sufficiently complex solution structure to ensure that all terms of the differential equations are exercised in the simulation. The MMS provides a straight forward and general procedure for generating such solutions. When used with systematic grid refinement studies, which are remarkably sensitive, the MMS provides strong code verification with a theorem-like quality. The MMS is first presented on simple 1-D examples. Manufactured solutions for more complex problems are then presented with sample results from grid convergence studies. (author)

A hybrid approach to quantify software reliability in nuclear safety systems

International Nuclear Information System (INIS)

Arun Babu, P.; Senthil Kumar, C.; Murali, N.

2012-01-01

Highlights: ► A novel method to quantify software reliability using software verification and mutation testing in nuclear safety systems. ► Contributing factors that influence software reliability estimate. ► Approach to help regulators verify the reliability of safety critical software system during software licensing process. -- Abstract: Technological advancements have led to the use of computer based systems in safety critical applications. As computer based systems are being introduced in nuclear power plants, effective and efficient methods are needed to ensure dependability and compliance to high reliability requirements of systems important to safety. Even after several years of research, quantification of software reliability remains controversial and unresolved issue. Also, existing approaches have assumptions and limitations, which are not acceptable for safety applications. This paper proposes a theoretical approach combining software verification and mutation testing to quantify the software reliability in nuclear safety systems. The theoretical results obtained suggest that the software reliability depends on three factors: the test adequacy, the amount of software verification carried out and the reusability of verified code in the software. The proposed approach may help regulators in licensing computer based safety systems in nuclear reactors.

Formal modelling and verification of interlocking systems featuring sequential release

DEFF Research Database (Denmark)

Vu, Linh Hong; Haxthausen, Anne Elisabeth; Peleska, Jan

2017-01-01

In this article, we present a method and an associated toolchain for the formal verification of the new Danish railway interlocking systems that are compatible with the European Train Control System (ETCS) Level 2. We have made a generic and reconfigurable model of the system behaviour and generic...... safety properties. This model accommodates sequential release - a feature in the new Danish interlocking systems. To verify the safety of an interlocking system, first a domain-specific description of interlocking configuration data is constructed and validated. Then the generic model and safety...

Procedure generation and verification

International Nuclear Information System (INIS)

Sheely, W.F.

1986-01-01

The Department of Energy has used Artificial Intelligence of ''AI'' concepts to develop two powerful new computer-based techniques to enhance safety in nuclear applications. The Procedure Generation System, and the Procedure Verification System, can be adapted to other commercial applications, such as a manufacturing plant. The Procedure Generation System can create a procedure to deal with the off-normal condition. The operator can then take correct actions on the system in minimal time. The Verification System evaluates the logic of the Procedure Generator's conclusions. This evaluation uses logic techniques totally independent of the Procedure Generator. The rapid, accurate generation and verification of corrective procedures can greatly reduce the human error, possible in a complex (stressful/high stress) situation

Dynamic Frames Based Verification Method for Concurrent Java Programs

NARCIS (Netherlands)

Mostowski, Wojciech

2016-01-01

In this paper we discuss a verification method for concurrent Java programs based on the concept of dynamic frames. We build on our earlier work that proposes a new, symbolic permission system for concurrent reasoning and we provide the following new contributions. First, we describe our approach

Specification and verification of the RTOS for plant protection systems

International Nuclear Information System (INIS)

Kim, Jin Hyun; Ahn, Young Ah; Lee, Su-Young; Choi, Jin Young; Lee, Na Young

2004-01-01

PLC is a computer system for instrumentation and control (I and C) systems such as control of machinery on factory assembly lines. control of machinery on factory assembly lines and Nucleare power plants. In nuclear power industry, systems is classified into 3 classes- Non-safety, safety-related and safety-critical up to integrity on system's using purpose. If PLC is used for controlling reactor in nuclear power plant, it should be identified as safety-critical. PLC has several I and C logics in software, including real-time operating system (RTOS). Hence, RTOS must be also proved that it is safe and reliable by various way and methods. In this paper, we apply formal methods to a development of RTOS for PLC in safety-critical level; Statecharts for specification and model checking for verification. In this paper, we give the results of applying formal methods to RTOS. (author)

KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Lee, Jang Soo; Eom, Heung Seop

1997-07-01

This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs.

KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

International Nuclear Information System (INIS)

Kim, Jang Yeol; Lee, Jang Soo; Eom, Heung Seop.

1997-07-01

This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs

Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

International Nuclear Information System (INIS)

Lee, A.G.; Wilkin, G.B.

1995-01-01

This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

Safety based on organisational learning (SOL) - Conceptual approach and verification of a method for event analysis

International Nuclear Information System (INIS)

Miller, R.; Wilpert, B.; Fahlbruch, B.

1999-01-01

This paper discusses a method for analysing safety-relevant events in NPP which is known as 'SOL', safety based on organisational learning. After discussion of the specific organisational and psychological problems examined in the event analysis, the analytic process using the SOL approach is explained as well as the required general setting. The SOL approach has been tested both with scientific experiments and from the practical perspective, by operators of NPPs and experts from other branches of industry. (orig./CB) [de

Verification and testing of the RTOS for safety-critical embedded systems

Energy Technology Data Exchange (ETDEWEB)

Lee, Na Young [Seoul National University, Seoul (Korea, Republic of); Kim, Jin Hyun; Choi, Jin Young [Korea University, Seoul (Korea, Republic of); Sung, Ah Young; Choi, Byung Ju [Ewha Womans University, Seoul (Korea, Republic of); Lee, Jang Soo [KAERI, Taejon (Korea, Republic of)

2003-07-01

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 code, the other is to find errors in the generated code. The former assumes that the verified model is correct, and the latter incorrect. Test data are generated separately to satisfy each purpose. After we test the RTOS software, we implement the test board embedded with the developed RTOS and the application software, which simulates the safety critical plant protection function. Testing to identify whether the reliability criteria is satisfied or not is also designed in this work. It results in that the developed RTOS software works well when it is embedded in the system.

Verification and testing of the RTOS for safety-critical embedded systems

International Nuclear Information System (INIS)

Lee, Na Young; Kim, Jin Hyun; Choi, Jin Young; Sung, Ah Young; Choi, Byung Ju; Lee, Jang Soo

2003-01-01

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 code, the other is to find errors in the generated code. The former assumes that the verified model is correct, and the latter incorrect. Test data are generated separately to satisfy each purpose. After we test the RTOS software, we implement the test board embedded with the developed RTOS and the application software, which simulates the safety critical plant protection function. Testing to identify whether the reliability criteria is satisfied or not is also designed in this work. It results in that the developed RTOS software works well when it is embedded in the system

The study of necessity of verification-methods for Depleted Uranium

International Nuclear Information System (INIS)

Park, J. B.; Ahn, S. H.; Ahn, G. H.; Chung, S. T.; Shin, J. S.

2006-01-01

ROK has tried to establish management system for depleted uranium from 2004, and ROK achieved some results in this field including management software, management skill, and the list of company using the nuclear material. But, the studies for the depleted uranium are insufficient exclude the studies of KAERI. In terms of SSAC, we have to study more about whether the depleted uranium is really dangerous material or not and how is the depleted uranium diverted to the nuclear weapon. The depleted uranium was controlled by the item counting in the national system for the small quantity nuclear material. We don't have unique technical methods to clarify the depleted uranium on-the-spot inspection not laboratory scale. Therefore, I would like to suggest of the necessity of the verification methods for depleted uranium. Furthermore, I would like to show you the methods of the verification of the depleted uranium in national system up to now

Safety Verification for Probabilistic Hybrid Systems

Czech Academy of Sciences Publication Activity Database

Zhang, J.; She, Z.; Ratschan, Stefan; Hermanns, H.; Hahn, E.M.

2012-01-01

Roč. 18, č. 6 (2012), s. 572-587 ISSN 0947-3580 R&D Projects: GA MŠk OC10048; GA ČR GC201/08/J020 Institutional research plan: CEZ:AV0Z10300504 Keywords : model checking * hybrid system s * formal verification Subject RIV: IN - Informatics, Computer Science Impact factor: 1.250, year: 2012

Accuracy verification methods theory and algorithms

CERN Document Server

Mali, Olli; Repin, Sergey

2014-01-01

The importance of accuracy verification methods was understood at the very beginning of the development of numerical analysis. Recent decades have seen a rapid growth of results related to adaptive numerical methods and a posteriori estimates. However, in this important area there often exists a noticeable gap between mathematicians creating the theory and researchers developing applied algorithms that could be used in engineering and scientific computations for guaranteed and efficient error control.   The goals of the book are to (1) give a transparent explanation of the underlying mathematical theory in a style accessible not only to advanced numerical analysts but also to engineers and students; (2) present detailed step-by-step algorithms that follow from a theory; (3) discuss their advantages and drawbacks, areas of applicability, give recommendations and examples.

Software V and V methods for a safety - grade programmable logic controller

International Nuclear Information System (INIS)

Jang Yeol Kim; Young Jun Lee; Kyung Ho Cha; Se Woo Cheon; Jang Soo Lee; Kee Choon Kwon

2006-01-01

This paper addresses the Verification and Validation(V and V) process and the methodology for an embedded real time software of a safety-grade Programmable Logic Controller(PLC). This safety- grade PLC is being developed as one of the Korean Nuclear Instrumentation and Control System (KNICS) projects. KNICS projects are developing a Reactor Protection System(RPS) and an Engineered Safety Feature-Component Control System(ESF-CCS) as well as a safety-grade PLC. The safety-grade PLC will be a major component that encomposes the RPS systems and the ESF-CCS systems as nuclear instruments and control equipment. This paper describes the V and V guidelines and procedures, V and V environment, V and V process and methodology, and the V and V tools in the KNICS projects. Specifically, it describes the real-time operating system V and V experience which corresponds to the requirement analysis phase, design phase and the implementation and testing phase of the software development life cycle. Main activities of the V and V for the PLC system software are a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and a software configuration management. The proposed V and V methodology satisfies the Standard Review Plan(SRP)/Branch Technical Position(BTP)-14 criteria for the safety software in nuclear power plants. The proposed V and V methodology is going to be used to verify the upcoming software life cycle in the KNICS projects. (author)

Verification of industrial x-ray machine: MINTs experience

International Nuclear Information System (INIS)

Aziz Amat; Saidi Rajab; Eesan Pasupathi; Saipo Bahari Abdul Ratan; Shaharudin Sayuti; Abd Nassir Ibrahim; Abd Razak Hamzah

2005-01-01

Radiation and electrical safety of the industrial x-ray equipment required to meet Atomic Energy Licensing Board(AELB) guidelines ( LEM/TEK/42 ) at the time of installation and subsequently a periodic verification should be ensured. The purpose of the guide is to explain the requirements employed in conducting the test on industrial x-ray apparatus and be certified in meeting with our local legislative and regulation. Verification is aimed to provide safety assurance information on electrical requirements and the minimum radiation exposure to the operator. This regulation is introduced on new models imported into the Malaysian market. Since June, 1997, Malaysian Institute for Nuclear Technology Research (MINT) has been approved by AELB to provide verification services to private company, government and corporate body throughout Malaysia. Early January 1997, AELB has made it mandatory that all x-ray equipment for industrial purpose (especially Industrial Radiography) must fulfill certain performance test based on the LEM/TEK/42 guidelines. MINT as the third party verification encourages user to improve maintenance of the equipment. MINT experiences in measuring the performance on intermittent and continuous duty rating single-phase industrial x-ray machine in the year 2004 indicated that all of irradiating apparatus tested pass the test and met the requirements of the guideline. From MINT record, 1997 to 2005 , three x-ray models did not meet the requirement and thus not allowed to be used unless the manufacturers willing to modify it to meet AELB requirement. This verification procedures on electrical and radiation safety on industrial x-ray has significantly improved the the maintenance cultures and safety awareness in the usage of x-ray apparatus in the industrial environment. (Author)

Development of the evaluation methods in reactor safety analyses and core characteristics

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

Nuclear cooperation targets global challenges. States back main pillars of the IAEA's work to strengthen nuclear safety, verification and technology transfer

International Nuclear Information System (INIS)

2000-01-01

States meeting at the 44th IAEA General Conference in Vienna have set a challenging agenda for international nuclear cooperation into the 21st century that targets issues of global safety, security, and sustainable development. They adopted resolutions endorsing the Agency's programmes for strengthening activities under its three main pillars of work - nuclear verification, safety, and technology - that are closely linked to major challenges before the world. The document presents the main actions taken during the conference

Spent Nuclear Fuel (SNF) Project Design Verification and Validation Process

International Nuclear Information System (INIS)

OLGUIN, L.J.

2000-01-01

This document provides a description of design verification and validation activities implemented by the Spent Nuclear Fuel (SNF) Project. During the execution of early design verification, a management assessment (Bergman, 1999) and external assessments on configuration management (Augustenburg, 1999) and testing (Loscoe, 2000) were conducted and identified potential uncertainties in the verification process. This led the SNF Chief Engineer to implement corrective actions to improve process and design products. This included Design Verification Reports (DVRs) for each subproject, validation assessments for testing, and verification of the safety function of systems and components identified in the Safety Equipment List to ensure that the design outputs were compliant with the SNF Technical Requirements. Although some activities are still in progress, the results of the DVR and associated validation assessments indicate that Project requirements for design verification are being effectively implemented. These results have been documented in subproject-specific technical documents (Table 2). Identified punch-list items are being dispositioned by the Project. As these remaining items are closed, the technical reports (Table 2) will be revised and reissued to document the results of this work

A feasible method for clinical delivery verification and dose reconstruction in tomotherapy

International Nuclear Information System (INIS)

Kapatoes, J.M.; Olivera, G.H.; Ruchala, K.J.; Smilowitz, J.B.; Reckwerdt, P.J.; Mackie, T.R.

2001-01-01

Delivery verification is the process in which the energy fluence delivered during a treatment is verified. This verified energy fluence can be used in conjunction with an image in the treatment position to reconstruct the full three-dimensional dose deposited. A method for delivery verification that utilizes a measured database of detector signal is described in this work. This database is a function of two parameters, radiological path-length and detector-to-phantom distance, both of which are computed from a CT image taken at the time of delivery. Such a database was generated and used to perform delivery verification and dose reconstruction. Two experiments were conducted: a simulated prostate delivery on an inhomogeneous abdominal phantom, and a nasopharyngeal delivery on a dog cadaver. For both cases, it was found that the verified fluence and dose results using the database approach agreed very well with those using previously developed and proven techniques. Delivery verification with a measured database and CT image at the time of treatment is an accurate procedure for tomotherapy. The database eliminates the need for any patient-specific, pre- or post-treatment measurements. Moreover, such an approach creates an opportunity for accurate, real-time delivery verification and dose reconstruction given fast image reconstruction and dose computation tools

Verification Survey of Uranium Mine Remediation

International Nuclear Information System (INIS)

Ron, Stager

2009-01-01

The Canadian Nuclear Safety Commission (CNSC) contracted an independent verification of an intensive gamma radiation survey conducted by a mining company to demonstrate that remediation of disturbed areas was complete. This site was the first of the recent mines being decommissioned in Canada and experience gained here may be applied to other mines being decommissioned in the future. The review included examination of the site-specific basis for clean-up criteria and ALARA as required by CNSC guidance. A paper review of the company report was conducted to determine if protocols were followed and that the summarized results could be independently reproduced. An independent verification survey was conducted on parts of the site and comparisons were made between gamma radiation measurements from the verification survey and the original company survey. Some aspects of data collection using rate meters linked to GPS data loggers are discussed as are aspects for data management and analyses methods required for the large amount of data collected during these surveys. Recommendations were made for implementation of future surveys and reporting the data from those surveys in order to ensure that remediation was complete. (authors)

Novel Verification Method for Timing Optimization Based on DPSO

Directory of Open Access Journals (Sweden)

Chuandong Chen

2018-01-01

Full Text Available Timing optimization for logic circuits is one of the key steps in logic synthesis. Extant research data are mainly proposed based on various intelligence algorithms. Hence, they are neither comparable with timing optimization data collected by the mainstream electronic design automation (EDA tool nor able to verify the superiority of intelligence algorithms to the EDA tool in terms of optimization ability. To address these shortcomings, a novel verification method is proposed in this study. First, a discrete particle swarm optimization (DPSO algorithm was applied to optimize the timing of the mixed polarity Reed-Muller (MPRM logic circuit. Second, the Design Compiler (DC algorithm was used to optimize the timing of the same MPRM logic circuit through special settings and constraints. Finally, the timing optimization results of the two algorithms were compared based on MCNC benchmark circuits. The timing optimization results obtained using DPSO are compared with those obtained from DC, and DPSO demonstrates an average reduction of 9.7% in the timing delays of critical paths for a number of MCNC benchmark circuits. The proposed verification method directly ascertains whether the intelligence algorithm has a better timing optimization ability than DC.

9 CFR 417.4 - Validation, Verification, Reassessment.

Science.gov (United States)

2010-01-01

.... 417.4 Section 417.4 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF... not have a HACCP plan because a hazard analysis has revealed no food safety hazards that are... ACT HAZARD ANALYSIS AND CRITICAL CONTROL POINT (HACCP) SYSTEMS § 417.4 Validation, Verification...

Programmable electronic system design & verification utilizing DFM

NARCIS (Netherlands)

Houtermans, M.J.M.; Apostolakis, G.E.; Brombacher, A.C.; Karydas, D.M.

2000-01-01

The objective of this paper is to demonstrate the use of the Dynamic Flowgraph Methodology (DIM) during the design and verification of programmable electronic safety-related systems. The safety system consists of hardware as well as software. This paper explains and demonstrates the use of DIM to

Formal verification method for nuclear I and C systems using ESDT and SMV in the software design phase

International Nuclear Information System (INIS)

Song, Myung Jun; Koo, Seo Ryong; Seong, Poong Hyun

2004-01-01

As PLCs are widely used in the digital I and C systems of nuclear power plants (NPPs), the safety of PLC software has become the most important consideration. Software safety is an important property for safety critical systems, especially those in aerospace, satellite and nuclear power plants, whose failure could result in danger to human life, property or environment. It is recently becoming more important due to the increase in the complexity and size of safety critical systems. This research proposes a method to perform effective verification tasks on the traceability analysis and software design evaluation in the software design phase. In order to perform the traceability analysis between a Software Requirements Specification (SRS) written in a natural language and a Software Design Specification (SDS) written in Function Block Diagram (FBD), this method uses extended-structured decision tables (ESDTs). ESDTs include information related to the traceability analysis from a text-based SRS and a FBD-based SDS, respectively. Through comparing with both ESDTs from an SRS and ESDTs from an SDS, the effective traceability analysis of both a text-based SRS and a FBD-based SDS can be achieved. For the software design evaluation, a model checking, which is mainly used to verify PLC programs formally, is used in this research. A FBD-style design specification is translated into input languages of the SMV by translation rules and then the FBD-style design specification can be formally analyzed using SMV. (author)

Verification and validation of software related to nuclear power plant control and instrumentation

International Nuclear Information System (INIS)

Wall, N.; Kossilov, A.

1994-01-01

There has always been significant concern with introduction of software in industry and the nuclear industry is no different from any other sector save its safety demands are some of the most onerous. The problems associated with software have led to the well documented difficulties in the introduction of computer based systems. An important area of concern with software in systems is the processes of Verification and Validation. One of the many activities the IAEA is currently engaged in is the preparation of a document on the process of verification and validation of software. The document follows the safety classification of IEC 1226 but includes software important to plant operation to establish three levels of assurance. The software that might be deployed on a plant was then identified as one of four types: new software, existing software for which full access to the code and documentation is possible, existing software of a proprietary nature and finally configurable software. The document attempts to identify the appropriate methods and tools for conducting the verification and validation processes. (author). 5 refs, 5 figs, 7 tabs

A method for online verification of adapted fields using an independent dose monitor

International Nuclear Information System (INIS)

Chang Jina; Norrlinger, Bernhard D.; Heaton, Robert K.; Jaffray, David A.; Cho, Young-Bin; Islam, Mohammad K.; Mahon, Robert

2013-01-01

Purpose: Clinical implementation of online adaptive radiotherapy requires generation of modified fields and a method of dosimetric verification in a short time. We present a method of treatment field modification to account for patient setup error, and an online method of verification using an independent monitoring system.Methods: The fields are modified by translating each multileaf collimator (MLC) defined aperture in the direction of the patient setup error, and magnifying to account for distance variation to the marked isocentre. A modified version of a previously reported online beam monitoring system, the integral quality monitoring (IQM) system, was investigated for validation of adapted fields. The system consists of a large area ion-chamber with a spatial gradient in electrode separation to provide a spatially sensitive signal for each beam segment, mounted below the MLC, and a calculation algorithm to predict the signal. IMRT plans of ten prostate patients have been modified in response to six randomly chosen setup errors in three orthogonal directions.Results: A total of approximately 49 beams for the modified fields were verified by the IQM system, of which 97% of measured IQM signal agree with the predicted value to within 2%.Conclusions: The modified IQM system was found to be suitable for online verification of adapted treatment fields

Method Verification Requirements for an Advanced Imaging System for Microbial Plate Count Enumeration.

Science.gov (United States)

Jones, David; Cundell, Tony

2018-01-01

The Growth Direct™ System that automates the incubation and reading of membrane filtration microbial counts on soybean-casein digest, Sabouraud dextrose, and R2A agar differs only from the traditional method in that micro-colonies on the membrane are counted using an advanced imaging system up to 50% earlier in the incubation. Based on the recommendations in USP Validation of New Microbiological Testing Methods , the system may be implemented in a microbiology laboratory after simple method verification and not a full method validation. LAY ABSTRACT: The Growth Direct™ System that automates the incubation and reading of microbial counts on membranes on solid agar differs only from the traditional method in that micro-colonies on the membrane are counted using an advanced imaging system up to 50% earlier in the incubation time. Based on the recommendations in USP Validation of New Microbiological Testing Methods , the system may be implemented in a microbiology laboratory after simple method verification and not a full method validation. © PDA, Inc. 2018.

Automated Installation Verification of COMSOL via LiveLink for MATLAB

International Nuclear Information System (INIS)

Crowell, Michael W

2015-01-01

Verifying that a local software installation performs as the developer intends is a potentially time-consuming but necessary step for nuclear safety-related codes. Automating this process not only saves time, but can increase reliability and scope of verification compared to ''hand'' comparisons. While COMSOL does not include automatic installation verification as many commercial codes do, it does provide tools such as LiveLink"T"M for MATLAB® and the COMSOL API for use with Java® through which the user can automate the process. Here we present a successful automated verification example of a local COMSOL 5.0 installation for nuclear safety-related calculations at the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR).

Automated Installation Verification of COMSOL via LiveLink for MATLAB

Energy Technology Data Exchange (ETDEWEB)

Crowell, Michael W [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-01-01

Verifying that a local software installation performs as the developer intends is a potentially time-consuming but necessary step for nuclear safety-related codes. Automating this process not only saves time, but can increase reliability and scope of verification compared to ‘hand’ comparisons. While COMSOL does not include automatic installation verification as many commercial codes do, it does provide tools such as LiveLink™ for MATLAB® and the COMSOL API for use with Java® through which the user can automate the process. Here we present a successful automated verification example of a local COMSOL 5.0 installation for nuclear safety-related calculations at the Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR).

Verification and Validation for Flight-Critical Systems (VVFCS)

Science.gov (United States)

Graves, Sharon S.; Jacobsen, Robert A.

2010-01-01

On March 31, 2009 a Request for Information (RFI) was issued by NASA s Aviation Safety Program to gather input on the subject of Verification and Validation (V & V) of Flight-Critical Systems. The responses were provided to NASA on or before April 24, 2009. The RFI asked for comments in three topic areas: Modeling and Validation of New Concepts for Vehicles and Operations; Verification of Complex Integrated and Distributed Systems; and Software Safety Assurance. There were a total of 34 responses to the RFI, representing a cross-section of academic (26%), small & large industry (47%) and government agency (27%).

Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

Energy Technology Data Exchange (ETDEWEB)

Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

Formal Verification of Continuous Systems

DEFF Research Database (Denmark)

Sloth, Christoffer

2012-01-01

and the verification procedures should be algorithmically synthesizable. Autonomous control plays an important role in many safety-critical systems. This implies that a malfunction in the control system can have catastrophic consequences, e.g., in space applications where a design flaw can result in large economic...... losses. Furthermore, a malfunction in the control system of a surgical robot may cause death of patients. The previous examples involve complex systems that are required to operate according to complex specifications. The systems cannot be formally verified by modern verification techniques, due...

Research on Linux Trusted Boot Method Based on Reverse Integrity Verification

Directory of Open Access Journals (Sweden)

Chenlin Huang

2016-01-01

Full Text Available Trusted computing aims to build a trusted computing environment for information systems with the help of secure hardware TPM, which has been proved to be an effective way against network security threats. However, the TPM chips are not yet widely deployed in most computing devices so far, thus limiting the applied scope of trusted computing technology. To solve the problem of lacking trusted hardware in existing computing platform, an alternative security hardware USBKey is introduced in this paper to simulate the basic functions of TPM and a new reverse USBKey-based integrity verification model is proposed to implement the reverse integrity verification of the operating system boot process, which can achieve the effect of trusted boot of the operating system in end systems without TPMs. A Linux operating system booting method based on reverse integrity verification is designed and implemented in this paper, with which the integrity of data and executable files in the operating system are verified and protected during the trusted boot process phase by phase. It implements the trusted boot of operation system without TPM and supports remote attestation of the platform. Enhanced by our method, the flexibility of the trusted computing technology is greatly improved and it is possible for trusted computing to be applied in large-scale computing environment.

V and V methods of a safety-critical software for a programmable logic controller

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Lee, Young Jun; Cha, Kyung Ho; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kong, Seung Ju [Korea Hydro and Nuclear Power Co., Ltd, Daejeon (Korea, Republic of)

2005-11-15

This paper addresses the Verification an Validation(V and V) process and the methodology for an embedded real time software of a safety-grade Programmable Logic Controller(PLC). This safety-grade PLC is being developed as one of the Korean Nuclear Instrumentation and Control System(KNICS) project KNICS projects are developing a Reactor Protection System(RPS) and an Engineered Safety Feature-Component Control System(ESF-CCS) as well as a safety-grade PLC. The safety-grade PLC will be a major component that encomposes the RPS systems and the ESF-CCS systems as nuclear instruments and control equipment. This paper describes the V and V guidelines an procedures, V and V environment, V and V process and methodology, and the V and V tools in the KNICS projects. Specifically, it describes the real-time operating system V and V experience which corresponds to the requirement analysis phase, design phase and the implementation and testing phase of the software development life cycle. Main activities of the V and V for the PLC system software are a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and a software configuration management. The proposed V and V methodology satisfies the Standard Review Plan(SRP)/Branch Technical Position(BTP)-14 criteria for the safety software in nuclear power plants. The proposed V and V methodology is going to be used to verify the upcoming software life cycle in the KNICS projects.

V and V methods of a safety-critical software for a programmable logic controller

International Nuclear Information System (INIS)

Kim, Jang Yeol; Lee, Young Jun; Cha, Kyung Ho; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon; Kong, Seung Ju

2005-01-01

This paper addresses the Verification an Validation(V and V) process and the methodology for an embedded real time software of a safety-grade Programmable Logic Controller(PLC). This safety-grade PLC is being developed as one of the Korean Nuclear Instrumentation and Control System(KNICS) project KNICS projects are developing a Reactor Protection System(RPS) and an Engineered Safety Feature-Component Control System(ESF-CCS) as well as a safety-grade PLC. The safety-grade PLC will be a major component that encomposes the RPS systems and the ESF-CCS systems as nuclear instruments and control equipment. This paper describes the V and V guidelines an procedures, V and V environment, V and V process and methodology, and the V and V tools in the KNICS projects. Specifically, it describes the real-time operating system V and V experience which corresponds to the requirement analysis phase, design phase and the implementation and testing phase of the software development life cycle. Main activities of the V and V for the PLC system software are a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and a software configuration management. The proposed V and V methodology satisfies the Standard Review Plan(SRP)/Branch Technical Position(BTP)-14 criteria for the safety software in nuclear power plants. The proposed V and V methodology is going to be used to verify the upcoming software life cycle in the KNICS projects

Validation and Verification of Future Integrated Safety-Critical Systems Operating under Off-Nominal Conditions

Science.gov (United States)

Belcastro, Christine M.

2010-01-01

Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents and reducing them will require a holistic integrated intervention capability. Future onboard integrated system technologies developed for preventing loss of vehicle control accidents must be able to assure safe operation under the associated off-nominal conditions. The transition of these technologies into the commercial fleet will require their extensive validation and verification (V and V) and ultimate certification. The V and V of complex integrated systems poses major nontrivial technical challenges particularly for safety-critical operation under highly off-nominal conditions associated with aircraft loss-of-control events. This paper summarizes the V and V problem and presents a proposed process that could be applied to complex integrated safety-critical systems developed for preventing aircraft loss-of-control accidents. A summary of recent research accomplishments in this effort is also provided.

A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods

International Nuclear Information System (INIS)

Benko, Pedro Luiz

1997-01-01

A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

Building quality into performance and safety assessment software

International Nuclear Information System (INIS)

Wojciechowski, L.C.

2011-01-01

Quality assurance is integrated throughout the development lifecycle for performance and safety assessment software. The software used in the performance and safety assessment of a Canadian deep geological repository (DGR) follows the CSA quality assurance standard CSA-N286.7 [1], Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants. Quality assurance activities in this standard include tasks such as verification and inspection; however, much more is involved in producing a quality software computer program. The types of errors found with different verification methods are described. The integrated quality process ensures that defects are found and corrected as early as possible. (author)

Study on safety classifications of software used in nuclear power plants and distinct applications of verification and validation activities in each class

International Nuclear Information System (INIS)

Kim, B. R.; Oh, S. H.; Hwang, H. S.; Kim, D. I.

2000-01-01

This paper describes the safety classification regarding instrumentation and control (I and C) systems and their software used in nuclear power plants, provides regulatory positions for software important to safety, and proposes verification and validation (V and V) activities applied differently in software classes which are important elements in ensuring software quality assurance. In other word, the I and C systems important to safety are classified into IC-1, IC-2, IC-3, and Non-IC and their software are classified into safety-critical, safety-related, and non-safety software. Based upon these safety classifications, the extent of software V and V activities in each class is differentiated each other. In addition, the paper presents that the software for use in I and C systems important to safety is divided into newly-developed and previously-developed software in terms of design and implementation, and provides the regulatory positions on each type of software

Statistical methods to correct for verification bias in diagnostic studies are inadequate when there are few false negatives: a simulation study

Directory of Open Access Journals (Sweden)

Vickers Andrew J

2008-11-01

Full Text Available Abstract Background A common feature of diagnostic research is that results for a diagnostic gold standard are available primarily for patients who are positive for the test under investigation. Data from such studies are subject to what has been termed "verification bias". We evaluated statistical methods for verification bias correction when there are few false negatives. Methods A simulation study was conducted of a screening study subject to verification bias. We compared estimates of the area-under-the-curve (AUC corrected for verification bias varying both the rate and mechanism of verification. Results In a single simulated data set, varying false negatives from 0 to 4 led to verification bias corrected AUCs ranging from 0.550 to 0.852. Excess variation associated with low numbers of false negatives was confirmed in simulation studies and by analyses of published studies that incorporated verification bias correction. The 2.5th – 97.5th centile range constituted as much as 60% of the possible range of AUCs for some simulations. Conclusion Screening programs are designed such that there are few false negatives. Standard statistical methods for verification bias correction are inadequate in this circumstance.

Young workers’ occupational safety knowledge creation and habits

OpenAIRE

Hejduk, Irena; Tomczyk, Przemysław

2015-01-01

The problem of young workers'safety culture is important because of the unfavorable demographic trend occurring in the European Union and determinants of competitiveness and innovativeness of the economy. The paper presents the concept and the importance of safety culture and goals of the research program, the aim of which is the construction and verification of the model based on the transfer of knowledge regarding the safety and methods of its implementation.Safety culture is a der...

Lightweight Methods for Effective Verification of Software Product Lines with Off-the-Shelf Tools

DEFF Research Database (Denmark)

Iosif-Lazar, Alexandru Florin

Certification is the process of assessing the quality of a product and whether it meets a set of requirements and adheres to functional and safety standards. I is often legally required to provide guarantee for human safety and to make the product available on the market. The certification process...... relies on objective evidence of quality, which is produced by using qualified and state-of-the-art tools and verification and validation techniques. Software product line (SPL) engineering distributes costs among similar products that are developed simultaneously. However, SPL certification faces major...... SPL reengineering projects that involve complex source code transformations. To facilitate product (re)certification, the transformation must preserve certain qualitative properties such as code structure and semantics—a difficult task due to the complexity of the transformation and because certain...

Compositional Verification of Interlocking Systems for Large Stations

DEFF Research Database (Denmark)

Fantechi, Alessandro; Haxthausen, Anne Elisabeth; Macedo, Hugo Daniel dos Santos

2017-01-01

-networks that are independent at some degree. At this regard, we study how the division of a complex network into sub-networks, using stub elements to abstract all the routes that are common between sub-networks, may still guarantee compositionality of verification of safety properties....... for networks of large size due to the exponential computation time and resources needed. Some recent attempts to address this challenge adopt a compositional approach, targeted to track layouts that are easily decomposable into sub-networks such that a route is almost fully contained in a sub......-network: in this way granting the access to a route is essentially a decision local to the sub-network, and the interfaces with the rest of the network easily abstract away less interesting details related to the external world. Following up on previous work, where we defined a compositional verification method...

Verification of reactor safety codes

International Nuclear Information System (INIS)

Murley, T.E.

1978-01-01

The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

A Method to Select Test Input Cases for Safety-critical Software

International Nuclear Information System (INIS)

Kim, Heeeun; Kang, Hyungook; Son, Hanseong

2013-01-01

This paper proposes a new testing methodology for effective and realistic quantification of RPS software failure probability. Software failure probability quantification is important factor in digital system safety assessment. In this study, the method for software test case generation is briefly described. The test cases generated by this method reflect the characteristics of safety-critical software and past inputs. Furthermore, the number of test cases can be reduced, but it is possible to perform exhaustive test. Aspect of software also can be reflected as failure data, so the final failure data can include the failure of software itself and external influences. Software reliability is generally accepted as the key factor in software quality since it quantifies software failures which can make a powerful system inoperative. In the KNITS (Korea Nuclear Instrumentation and Control Systems) project, the software for the fully digitalized reactor protection system (RPS) was developed under a strict procedure including unit testing and coverage measurement. Black box testing is one type of Verification and validation (V and V), in which given input values are entered and the resulting output values are compared against the expected output values. Programmable logic controllers (PLCs) were used in implementing critical systems and function block diagram (FBD) is a commonly used implementation language for PLC

Discussion on verification criterion and method of human factors engineering for nuclear power plant controller

International Nuclear Information System (INIS)

Yang Hualong; Liu Yanzi; Jia Ming; Huang Weijun

2014-01-01

In order to prevent or reduce human error and ensure the safe operation of nuclear power plants, control device should be verified from the perspective of human factors engineering (HFE). The domestic and international human factors engineering guidelines about nuclear power plant controller were considered, the verification criterion and method of human factors engineering for nuclear power plant controller were discussed and the application examples were provided for reference in this paper. The results show that the appropriate verification criterion and method should be selected to ensure the objectivity and accuracy of the conclusion. (authors)

State Token Petri Net modeling method for formal verification of computerized procedure including operator's interruptions of procedure execution flow

International Nuclear Information System (INIS)

Kim, Yun Goo; Seong, Poong Hyun

2012-01-01

The Computerized Procedure System (CPS) is one of the primary operating support systems in the digital Main Control Room. The CPS displays procedure on the computer screen in the form of a flow chart, and displays plant operating information along with procedure instructions. It also supports operator decision making by providing a system decision. A procedure flow should be correct and reliable, as an error would lead to operator misjudgement and inadequate control. In this paper we present a modeling for the CPS that enables formal verification based on Petri nets. The proposed State Token Petri Nets (STPN) also support modeling of a procedure flow that has various interruptions by the operator, according to the plant condition. STPN modeling is compared with Coloured Petri net when they are applied to Emergency Operating Computerized Procedure. A converting program for Computerized Procedure (CP) to STPN has been also developed. The formal verification and validation methods of CP with STPN increase the safety of a nuclear power plant and provide digital quality assurance means that are needed when the role and function of the CPS is increasing.

Qualification of safety-critical software for digital reactor safety system in nuclear power plants

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo

2013-01-01

This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)

Safety assessment and verification for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

Safety assessment and verification for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

General-purpose heat source safety verification test series: SVT-11 through SVT-13

International Nuclear Information System (INIS)

George, T.G.; Pavone, D.

1986-05-01

The General-Purpose Heat Source (GPHS) is a modular component of the radioisotope thermoelectric generator that will provide power for the Galileo and Ulysses (formerly ISPM) space missions. The GPHS provides power by transmitting the heat of 238 Pu α-decay to an array of thermoelectric elements. Because the possibility of an orbital abort always exists, the heat source was designed and constructed to minimize plutonia release in any accident environment. The Safety Verification Test (SVT) series was formulated to evaluate the effectiveness of GPHS plutonia containment after atmospheric reentry and Earth impact. The first two reports (covering SVT-1 through SVT-10) described the results of flat, side-on, and angular module impacts against steel targets at 54 m/s. This report describes flat-on module impacts against concrete and granite targets, at velocities equivalent to or higher than previous SVTs

HDL to verification logic translator

Science.gov (United States)

Gambles, J. W.; Windley, P. J.

1992-01-01

The increasingly higher number of transistors possible in VLSI circuits compounds the difficulty in insuring correct designs. As the number of possible test cases required to exhaustively simulate a circuit design explodes, a better method is required to confirm the absence of design faults. Formal verification methods provide a way to prove, using logic, that a circuit structure correctly implements its specification. Before verification is accepted by VLSI design engineers, the stand alone verification tools that are in use in the research community must be integrated with the CAD tools used by the designers. One problem facing the acceptance of formal verification into circuit design methodology is that the structural circuit descriptions used by the designers are not appropriate for verification work and those required for verification lack some of the features needed for design. We offer a solution to this dilemma: an automatic translation from the designers' HDL models into definitions for the higher-ordered logic (HOL) verification system. The translated definitions become the low level basis of circuit verification which in turn increases the designer's confidence in the correctness of higher level behavioral models.

Formal model-based development for safety-critical embedded software

International Nuclear Information System (INIS)

Kim, Jin Hyun; Choi, Jin Young

2005-01-01

Safety-critical embedded software for nuclear I and C system is developed under the safety and reliability regulation. Programmable logic controller(PLC) is a computer system for instrumentation and control (I and C) system of nuclear power plants. PLC consists of various I and C logics in software, including real-time operating system (RTOS). Hence, errors related with RTOS should be detected and eliminated in development processes. Practically, the verification and validation for errors in RTOS is performed in test procedure, in which a lot of tasks for testing are embedded in RTOS and are running under a test environments. But the test process can not be enough to guarantee the safety and reliability of RTOS. Therefore, in this paper, we introduce to applying formal methods with the development of software for the PLC. We particularity apply formal methods to a development of RTOS for PLC, which is a safety critical level. In this development, we use the state charts of I-Logix to specify and verification and model checking to verify the specification

Formal model-based development for safety-critical embedded software

Energy Technology Data Exchange (ETDEWEB)

Kim, Jin Hyun; Choi, Jin Young [Korea University, seoul (Korea, Republic of)

2005-11-15

Safety-critical embedded software for nuclear I and C system is developed under the safety and reliability regulation. Programmable logic controller(PLC) is a computer system for instrumentation and control (I and C) system of nuclear power plants. PLC consists of various I and C logics in software, including real-time operating system (RTOS). Hence, errors related with RTOS should be detected and eliminated in development processes. Practically, the verification and validation for errors in RTOS is performed in test procedure, in which a lot of tasks for testing are embedded in RTOS and are running under a test environments. But the test process can not be enough to guarantee the safety and reliability of RTOS. Therefore, in this paper, we introduce to applying formal methods with the development of software for the PLC. We particularity apply formal methods to a development of RTOS for PLC, which is a safety critical level. In this development, we use the state charts of I-Logix to specify and verification and model checking to verify the specification.

78 FR 28812 - Energy Efficiency Program for Industrial Equipment: Petition of UL Verification Services Inc. for...

Science.gov (United States)

2013-05-16

... are engineers. UL today is comprised of five businesses, Product Safety, Verification Services, Life..., Director--Global Technical Research, UL Verification Services. Subscribed and sworn to before me this 20... (431.447(c)(4)) General Personnel Overview UL is a global independent safety science company with more...

Safety training for working youth: Methods used versus methods wanted.

Science.gov (United States)

Zierold, Kristina M

2016-04-07

Safety training is promoted as a tool to prevent workplace injury; however, little is known about the safety training experiences young workers get on-the-job. Furthermore, nothing is known about what methods they think would be the most helpful for learning about safe work practices. To compare safety training methods teens get on the job to those safety training methods teens think would be the best for learning workplace safety, focusing on age differences. A cross-sectional survey was administered to students in two large high schools in spring 2011. Seventy percent of working youth received safety training. The top training methods that youth reported getting at work were safety videos (42%), safety lectures (25%), and safety posters/signs (22%). In comparison to the safety training methods used, the top methods youth wanted included videos (54%), hands-on (47%), and on-the-job demonstrations (34%). This study demonstrated that there were differences in training methods that youth wanted by age; with older youth seemingly wanting more independent methods of training and younger teens wanting more involvement. Results indicate that youth want methods of safety training that are different from what they are getting on the job. The differences in methods wanted by age may aid in developing training programs appropriate for the developmental level of working youth.

Tensit - a novel probabilistic simulation tool for safety assessments. Tests and verifications using biosphere models

Energy Technology Data Exchange (ETDEWEB)

Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

2004-06-01

This report documents the verification of a new simulation tool for dose assessment put together in a package under the name Tensit (Technical Nuclide Simulation Tool). The tool is developed to solve differential equation systems describing transport and decay of radionuclides. It is capable of handling both deterministic and probabilistic simulations. The verifications undertaken shows good results. Exceptions exist only where the reference results are unclear. Tensit utilise and connects two separate commercial softwares. The equation solving capability is derived from the Matlab/Simulink software environment to which Tensit adds a library of interconnectable building blocks. Probabilistic simulations are provided through a statistical software named at{sub R}isk that communicates with Matlab/Simulink. More information about these softwares can be found at www.palisade.com and www.mathworks.com. The underlying intention of developing this new tool has been to make available a cost efficient and easy to use means for advanced dose assessment simulations. The mentioned benefits are gained both through the graphical user interface provided by Simulink and at{sub R}isk, and the use of numerical equation solving routines in Matlab. To verify Tensit's numerical correctness, an implementation was done of the biosphere modules for dose assessments used in the earlier safety assessment project SR 97. Acquired probabilistic results for deterministic as well as probabilistic simulations have been compared with documented values. Additional verification has been made both with another simulation tool named AMBER and also against the international test case from PSACOIN named Level 1B. This report documents the models used for verification with equations and parameter values so that the results can be recreated. For a background and a more detailed description of the underlying processes in the models, the reader is referred to the original references. Finally, in the

Tensit - a novel probabilistic simulation tool for safety assessments. Tests and verifications using biosphere models

International Nuclear Information System (INIS)

Jones, Jakob; Vahlund, Fredrik; Kautsky, Ulrik

2004-06-01

This report documents the verification of a new simulation tool for dose assessment put together in a package under the name Tensit (Technical Nuclide Simulation Tool). The tool is developed to solve differential equation systems describing transport and decay of radionuclides. It is capable of handling both deterministic and probabilistic simulations. The verifications undertaken shows good results. Exceptions exist only where the reference results are unclear. Tensit utilise and connects two separate commercial softwares. The equation solving capability is derived from the Matlab/Simulink software environment to which Tensit adds a library of interconnectable building blocks. Probabilistic simulations are provided through a statistical software named at R isk that communicates with Matlab/Simulink. More information about these softwares can be found at www.palisade.com and www.mathworks.com. The underlying intention of developing this new tool has been to make available a cost efficient and easy to use means for advanced dose assessment simulations. The mentioned benefits are gained both through the graphical user interface provided by Simulink and at R isk, and the use of numerical equation solving routines in Matlab. To verify Tensit's numerical correctness, an implementation was done of the biosphere modules for dose assessments used in the earlier safety assessment project SR 97. Acquired probabilistic results for deterministic as well as probabilistic simulations have been compared with documented values. Additional verification has been made both with another simulation tool named AMBER and also against the international test case from PSACOIN named Level 1B. This report documents the models used for verification with equations and parameter values so that the results can be recreated. For a background and a more detailed description of the underlying processes in the models, the reader is referred to the original references. Finally, in the perspective of

VBMC: a formal verification tool for VHDL programs

International Nuclear Information System (INIS)

Ajith, K.J.; Bhattacharjee, A.K.

2014-01-01

The design of Control and Instrumentation (C and I) systems used in safety critical applications such as nuclear power plants involves partitioning of the overall system functionality into subparts and implementing each subpart in hardware and/or software as appropriate. With increasing use of programmable devices like FPGA, the hardware subsystems are often implemented in Hardware Description Languages (HDL) like VHDL. Since the functional bugs in such hardware subsystems used in safety critical C and I systems have disastrous consequences, it is important to use rigorous reasoning to verify the functionalities of the HDL models. This paper describes an indigenously developed software tool named VBMC (VHDL Bounded Model Checker) for mathematically proving/refuting functional properties of hardware designs described in VHDL. VBMC accepts hardware design as VHDL program file, functional property in PSL, and verification bound (number of cycles of operation) as inputs. It either reports that the design satisfies the functional property for the given verification bound or generates a counter example providing the reason of violation. In case of satisfaction, the proof holds good for the verification bound. VBMC has been used for the functional verification of FPGA based intelligent I/O boards developed at Reactor Control Division, BARC. (author)

VBMC: a formal verification tool for VHDL program

International Nuclear Information System (INIS)

Ajith, K.J.; Bhattacharjee, A.K.

2014-08-01

The design of Control and Instrumentation (C and I) systems used in safety critical applications such as nuclear power plants involves partitioning of the overall system functionality into sub-parts and implementing each sub-part in hardware and/or software as appropriate. With increasing use of programmable devices like FPGA, the hardware subsystems are often implemented in Hardware Description Languages (HDL) like VHDL. Since the functional bugs in such hardware subsystems used in safety critical C and I systems have serious consequences, it is important to use rigorous reasoning to verify the functionalities of the HDL models. This report describes the design of a software tool named VBMC (VHDL Bounded Model Checker). The capability of this tool is in proving/refuting functional properties of hardware designs described in VHDL. VBMC accepts design as a VHDL program file, functional property in PSL, and verification bound (number of cycles of operation) as inputs. It either reports that the design satisfies the functional property for the given verification bound or generates a counterexample providing the reason of violation. In case of satisfaction, the proof holds good for the verification bound. VBMC has been used for the functional verification of FPGA based intelligent I/O boards developed at Reactor Control Division, BARC. (author)

Safety technical considerations on the 2012 periodic safety verification of the Beznau nuclear power plant

International Nuclear Information System (INIS)

2016-12-01

According to nuclear legislation, the owner of an operational license for a nuclear power plant has to provide a periodic safety verification (PSU) every 10 years. The 'North Eastern Power Plants' company (NOK), today AXPO Power AG already performed such a PSU for the Beznau-2 nuclear reactor block (KKB2) in 2002. The Beznau-1 nuclear reactor block (KKB1) received its definitive operational license in October 1970, after test operation during 7 months. After the license for test operation received on July 16 th , 1971, the operational license of KKB2 was renewed several times, each time for a certain period of validity. In 1991, NOK requested a definitive operational license for KKB2, but in 1994 the Swiss Federal Council lengthened the license for only 10 years. Moreover, it laid down that NOK has to periodically report on the safety of the facility. With its letter of August 23 rd , 1998, the Federal Office of Energy defined the documents to be produced for the PSU. The extent of the PSU was defined in such a way that many documents concern the whole power plant, i.e. both nuclear reactor blocks. On December 3 rd , 2004, the Swiss Federal Council granted KKB2 an operational license of limited validity. The present report reviews the 2012 PSU, which covers the time interval from January 1 st , 2002, to December 31 st , 2011, from the point of view of safety. It contains documents for the evaluation of both reactor blocks at KKB. The Beznau interim storage pool was also taken into consideration; it is situated on the KKB site, but, according to a decision of the Swiss Federal Council of May 23 rd , 1991, it has an independent operational license. The evaluation of ageing surveillance takes the whole operational period of the facility into account, i.e. the ageing mechanisms acting as from the beginning of the operation. Moreover, important developments that occurred after the surveillance time interval have been taken into account, especially the status

SIMMER-III code-verification. Phase 1

International Nuclear Information System (INIS)

Maschek, W.

1996-05-01

SIMMER-III is a computer code to investigate core disruptive accidents in liquid metal fast reactors but should also be used to investigate safety related problems in other types of advanced reactors. The code is developed by PNC with cooperation of the European partners FZK, CEA and AEA-T. SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-, time-, and energy-dependent neutron dynamics model. In order to model complex flow situations in a postulated disrupting core, mass and energy conservation equations are solved for 27 density components and 16 energy components, respectively. Three velocity fields (two liquid and one vapor) are modeled to simulate the relative motion of different fluid components. An additional static field takes into account the structures available in a reactor (pins, hexans, vessel structures, internal structures etc.). The neutronics is based on the discrete ordinate method (S N method) coupled into a quasistatic dynamic model. The code assessment and verification of the fluid dynamic/thermohydraulic parts of the code is performed in several steps in a joint effort of all partners. The results of the FZK contributions to the first assessment and verification phase is reported. (orig.) [de

A simple method for validation and verification of pipettes mounted on automated liquid handlers

DEFF Research Database (Denmark)

Stangegaard, Michael; Hansen, Anders Johannes; Frøslev, Tobias Guldberg

 We have implemented a simple method for validation and verification of the performance of pipettes mounted on automated liquid handlers as necessary for laboratories accredited under ISO 17025. An 8-step serial dilution of Orange G was prepared in quadruplicates in a flat bottom 96-well microtit...... available. In conclusion, we have set up a simple solution for the continuous validation of automated liquid handlers used for accredited work. The method is cheap, simple and easy to use for aqueous solutions but requires a spectrophotometer that can read microtiter plates....... We have implemented a simple method for validation and verification of the performance of pipettes mounted on automated liquid handlers as necessary for laboratories accredited under ISO 17025. An 8-step serial dilution of Orange G was prepared in quadruplicates in a flat bottom 96-well microtiter...

Methods and procedures for the verification and validation of artificial neural networks

CERN Document Server

Taylor, Brian J

2006-01-01

Neural networks are members of a class of software that have the potential to enable intelligent computational systems capable of simulating characteristics of biological thinking and learning. This volume introduces some of the methods and techniques used for the verification and validation of neural networks and adaptive systems.

Contribution of NDT to the safety of pressurized components in power stations

International Nuclear Information System (INIS)

Mletzko, U.; Maier, H.J.

1994-01-01

In the eyes on the MPA Stuttgart, the nondestructive testing has a very high weight relating to the safety of pressure components in power stations (concept of basis safety). In this connection, the performance verification by NDT has a special significance. A qualification of NDT-techniques can be, indeed, executed in the initial stage at test bodies with artificial faults, known with respect to position, size and type. Even theoretical (modelling) considerations can be integrated. For a performance verification in a closer sense however, this is not sufficient. The performance verification should be effected for the overall system, composed of hardware, software and examination personnel at components having the scale of 1:1 (Full Scale) under realistic boundary conditions and given times. The components must have natural or quasi-natural faults in a certain quality. The informative value of performance verifications is considerably increased, when executed as authenitic dry runs, and when the fault state is subsequently verified by destructive (metallographical) methods. (orig.) [de

General safety considerations

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

General safety considerations

International Nuclear Information System (INIS)

2001-01-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

General safety considerations

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-09-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

General safety considerations

International Nuclear Information System (INIS)

1998-01-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

On-power verification of the dynamic response of self-powered in-core detectors

International Nuclear Information System (INIS)

Serdula, K.; Beaudet, M.

1996-01-01

Self-powered in-core detectors are used for on-line safety and regulation purposes in CANDU reactors. Such applications require use of detectors whose response is primarily prompt to changes in flux. In-service verification of the detectors' response is required to ensure significant degradation in performance has not occurred during long-term operation. Changes in the detector characteristics occur due to nuclear interactions and failures. Present verification requires significant station resources and disrupts power production. Use of the 'noise' in the detector signal is being investigated as an alternative to assess the dynamic response of the detectors during long-term operation. Measurements of reference 'signatures' were obtained from replacement shutdown system detectors. Results show 'noise' measurements are a promising alternative to the current verification method. Identification of changes in the detector response function assist in accurate diagnosis and prognosis of changes in detector signals due to process changes. (author)

9 CFR 417.8 - Agency verification.

Science.gov (United States)

2010-01-01

....8 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE... ANALYSIS AND CRITICAL CONTROL POINT (HACCP) SYSTEMS § 417.8 Agency verification. FSIS will verify the... plan or system; (f) Direct observation or measurement at a CCP; (g) Sample collection and analysis to...

Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

Energy Technology Data Exchange (ETDEWEB)

Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-06-01

The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

General safety aspects

International Nuclear Information System (INIS)

1998-01-01

In this part next aspects are described: (1) Priority to safety; (2) Financial and human resources;; (3) Human factor; (4) Operator's quality assurance system; (5) Safety assessment and Verification; (6) Radiation protection and (7) Emergency preparedness

Method of V ampersand V for safety-critical software in NPPs

International Nuclear Information System (INIS)

Kim, Jang-Yeol; Lee, Jang-Soo; Kwon, Kee-Choon

1997-01-01

Safety-critical software is software used in systems in which a failure could affect personal or equipment safety or result in large financial or social loss. Examples of systems using safety-critical software are systems such as plant protection systems in nuclear power plants (NPPs), process control systems in chemical plants, and medical instruments such as the Therac-25 medical accelerator. This paper presents verification and validation (V ampersand V) methodology for safety-critical software in NPP safety systems. In addition, it addresses issues related to NPP safety systems, such as independence parameters, software safety analysis (SSA) concepts, commercial off-the-shelf (COTS) software evaluation criteria, and interrelationships among software and system assurance organizations. It includes the concepts of existing industrial standards on software V ampersand V, Institute of Electrical and Electronics Engineers (IEEE) Standards 1012 and 1059. This safety-critical software V ampersand V methodology covers V ampersand V scope, a regulatory framework as part of its acceptance criteria, V ampersand V activities and task entrance and exit criteria, reviews and audits, testing and quality assurance records of V ampersand V material, configuration management activities related to V ampersand V, and software V ampersand V (SVV) plan (SVVP) production

Investigation of a Verification and Validation Tool with a Turbofan Aircraft Engine Application

Science.gov (United States)

Uth, Peter; Narang-Siddarth, Anshu; Wong, Edmond

2018-01-01

The development of more advanced control architectures for turbofan aircraft engines can yield gains in performance and efficiency over the lifetime of an engine. However, the implementation of these increasingly complex controllers is contingent on their ability to provide safe, reliable engine operation. Therefore, having the means to verify the safety of new control algorithms is crucial. As a step towards this goal, CoCoSim, a publicly available verification tool for Simulink, is used to analyze C-MAPSS40k, a 40,000 lbf class turbo-fan engine model developed at NASA for testing new control algorithms. Due to current limitations of the verification software, several modifications are made to C-MAPSS40k to achieve compatibility with CoCoSim. Some of these modifications sacrifice fidelity to the original model. Several safety and performance requirements typical for turbofan engines are identified and constructed into a verification framework. Preliminary results using an industry standard baseline controller for these requirements are presented. While verification capabilities are demonstrated, a truly comprehensive analysis will require further development of the verification tool.

Verification of Ceramic Structures

Science.gov (United States)

Behar-Lafenetre, Stephanie; Cornillon, Laurence; Rancurel, Michael; De Graaf, Dennis; Hartmann, Peter; Coe, Graham; Laine, Benoit

2012-07-01

In the framework of the “Mechanical Design and Verification Methodologies for Ceramic Structures” contract [1] awarded by ESA, Thales Alenia Space has investigated literature and practices in affiliated industries to propose a methodological guideline for verification of ceramic spacecraft and instrument structures. It has been written in order to be applicable to most types of ceramic or glass-ceramic materials - typically Cesic®, HBCesic®, Silicon Nitride, Silicon Carbide and ZERODUR®. The proposed guideline describes the activities to be performed at material level in order to cover all the specific aspects of ceramics (Weibull distribution, brittle behaviour, sub-critical crack growth). Elementary tests and their post-processing methods are described, and recommendations for optimization of the test plan are given in order to have a consistent database. The application of this method is shown on an example in a dedicated article [7]. Then the verification activities to be performed at system level are described. This includes classical verification activities based on relevant standard (ECSS Verification [4]), plus specific analytical, testing and inspection features. The analysis methodology takes into account the specific behaviour of ceramic materials, especially the statistical distribution of failures (Weibull) and the method to transfer it from elementary data to a full-scale structure. The demonstration of the efficiency of this method is described in a dedicated article [8]. The verification is completed by classical full-scale testing activities. Indications about proof testing, case of use and implementation are given and specific inspection and protection measures are described. These additional activities are necessary to ensure the required reliability. The aim of the guideline is to describe how to reach the same reliability level as for structures made of more classical materials (metals, composites).

Development of synchronized control method for shaking table with booster device. Verification of the capabilities based on both real facility and numerical simulator

International Nuclear Information System (INIS)

Kajii, Shin-ichirou; Yasuda, Chiaki; Yamashita, Toshio; Abe, Hiroshi; Kanki, Hiroshi

2004-01-01

In the seismic design of nuclear power plant, it is recently considered to use probability method in a addition to certainty method. The former method is called Seismic Probability Safety Assessment (Seismic PSA). In case of seismic PSA for some components of a nuclear power plant using a shaking table, it is necessary for some limited conditions with high level of accelerations such as actual conditions. However, it might be difficult to achieve the test conditions that a current shaking table based on hydraulic power system is intended for the test facility. Therefore, we have been planning out a test method in which both a current and another shaking table called a booster device are applied. This paper describes the verification test of a synchronized control between a current shaking table and a booster device. (author)

MDEP Generic Common Position No DICWG-03. Common position on verification and validation throughout the life cycle of digital safety systems

International Nuclear Information System (INIS)

2013-01-01

Verification and validation (V and V) is essential throughout the life cycle of nuclear power plant safety systems. This common position applies to V and V activities for digital safety systems throughout their life cycles. This encompasses both the software and hardware of such systems. The Digital Instrumentation and Controls Working Group (DICWG) has agreed that a common position on this topic is warranted given the use of Digital I and C in new reactor designs, its safety implications, and the need to develop a common understanding from the perspectives of regulatory authorities. This action follows the DICWG examination of the regulatory requirements of the participating members and of relevant industry standards and IAEA documents. The DICWG proposes a common position based on its recent experience with the new reactor application reviews and operating plant issues

Cuban experience in verification of the execution of the safety requirements during the transport of radioactive materials

International Nuclear Information System (INIS)

Quevedo Garcia, J.R.; Lopez Forteza, Y.

2001-01-01

The Cuban Regulatory Authority has paid special attention to the verification of the execution of the safety requirements during the transport of radioactive material in the country. With this purpose, the Authority has followed a consequent policy based on supplementary demands to those collections in the juridical mark settled down in 1987 in the sphere of transport of radioactive substances. In the work the technical approaches are exposed kept in mind when establishing the one referred politics, the current situation is characterized, the results are evaluated obtained in correspondence with the pursued objectives and the essential aspects are exposed to keep in mind for the adopted politics ulterior development. (author)

Scalable Techniques for Formal Verification

CERN Document Server

Ray, Sandip

2010-01-01

This book presents state-of-the-art approaches to formal verification techniques to seamlessly integrate different formal verification methods within a single logical foundation. It should benefit researchers and practitioners looking to get a broad overview of the spectrum of formal verification techniques, as well as approaches to combining such techniques within a single framework. Coverage includes a range of case studies showing how such combination is fruitful in developing a scalable verification methodology for industrial designs. This book outlines both theoretical and practical issue

Content-based Image Hiding Method for Secure Network Biometric Verification

Directory of Open Access Journals (Sweden)

Xiangjiu Che

2011-08-01

Full Text Available For secure biometric verification, most existing methods embed biometric information directly into the cover image, but content correlation analysis between the biometric image and the cover image is often ignored. In this paper, we propose a novel biometric image hiding approach based on the content correlation analysis to protect the network-based transmitted image. By using principal component analysis (PCA, the content correlation between the biometric image and the cover image is firstly analyzed. Then based on particle swarm optimization (PSO algorithm, some regions of the cover image are selected to represent the biometric image, in which the cover image can carry partial content of the biometric image. As a result of the correlation analysis, the unrepresented part of the biometric image is embedded into the cover image by using the discrete wavelet transform (DWT. Combined with human visual system (HVS model, this approach makes the hiding result perceptually invisible. The extensive experimental results demonstrate that the proposed hiding approach is robust against some common frequency and geometric attacks; it also provides an effective protection for the secure biometric verification.

Model checking of safety-critical software in the nuclear engineering domain

International Nuclear Information System (INIS)

Lahtinen, J.; Valkonen, J.; Björkman, K.; Frits, J.; Niemelä, I.; Heljanko, K.

2012-01-01

Instrumentation and control (I and C) systems play a vital role in the operation of safety-critical processes. Digital programmable logic controllers (PLC) enable sophisticated control tasks which sets high requirements for system validation and verification methods. Testing and simulation have an important role in the overall verification of a system but are not suitable for comprehensive evaluation because only a limited number of system behaviors can be analyzed due to time limitations. Testing is also performed too late in the development lifecycle and thus the correction of design errors is expensive. This paper discusses the role of formal methods in software development in the area of nuclear engineering. It puts forward model checking, a computer-aided formal method for verifying the correctness of a system design model, as a promising approach to system verification. The main contribution of the paper is the development of systematic methodology for modeling safety critical systems in the nuclear domain. Two case studies are reviewed, in which we have found errors that were previously not detected. We also discuss the actions that should be taken in order to increase confidence in the model checking process.

Status of safety issues at licensed power plants: TMI Action Plan requirements; unresolved safety issues; generic safety issues; other multiplant action issues

International Nuclear Information System (INIS)

1993-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. This third annual NUREG report, Supplement 3, presents updated information as of September 30, 1993. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

Risk analysis methods: their importance for safety assessment of practices using radiation

International Nuclear Information System (INIS)

Dumenigo, C; Vilaragut, J.J.; Ferro, R.; Guillen, A.; Ramirez, M.L.; Ortiz Lopez, P.; Rodriguez, M.; McDonnell, J.D.; Papadopulos, S.; Pereira, P.P.; Goncalvez, M.; Morales, J.; Larrinaga, E.; Lopez Morones, R.; Sanchez, R.; Delgado, J.M.; Sanchez, C.; Somoano, F.

2008-01-01

Radiation safety has been based for many years on verification of compliance with regulatory requirements, codes of practice and international standards, which can be considered prescriptive methods. Accident analyses have been published, lessons have been learned and safety assessments have incorporated the need to check whether a facility is ready to avoid accidents similar to the reported ones. These approaches can be also called 'reactive methods'. They have in common the fundamental limitation of being restricted to reported experience, but do not take into account other potential events, which were never published or never happened, i.e. latent risks. Moreover, they focus on accident sequences with major consequences and low probability but may not pay enough attention to other sequences leading to lower, but still significant consequences with higher probability. More proactive approaches are, therefore, needed, to assess risk in radiation facilities. They aim at identifying all potential equipment faults and human error, which can lead to predefined unwanted consequences and are based on the general risk equation: Risk = Probability of occurrence of an accidental sequence * magnitude of the consequences. In this work, a review is given of the experience obtained by the countries of the Ibero American Forum of Nuclear and Radiation Safety Regulatory Organizations, by applying proactive methods to radiotherapy practice. In particular, probabilistic safety assessment (PSA) used for external beam treatments with linear electron accelerators and two studies, on cobalt 60 therapy and brachytherapy using the risk-matrix approach are presented. The work has identified event sequences, their likelihood of occurrence, the consequences, the efficiency of interlocks and control checks and the global importance in terms of overall risk, to facilitate decision making and implementation of preventive measures. A comparison is presented of advantages and limitations of

Status of safety issues at licensed power plants: TMI action plan requirements, unresolved safety issues, generic safety issues

International Nuclear Information System (INIS)

1991-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program was established whereby an annual NUREG report would be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was compiled and reported in three NUREG volumes. Volume 1, published in March 1991, addressed the status of of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). This annual NUREG report combines these volumes into a single report and provides updated information as of September 30, 1991. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. This report is to provide a comprehensive description of the implementation and verification status of TMI Action Plan Requirements, safety issues designated as USIs, and GSIs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

International Nuclear Information System (INIS)

Lee, A.G.; Wilkin, G.B.

1996-03-01

During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

A method of knowledge base verification for nuclear power plant expert systems using extended Petri Nets

International Nuclear Information System (INIS)

Kwon, I. W.; Seong, P. H.

1996-01-01

The adoption of expert systems mainly as operator supporting systems is becoming increasingly popular as the control algorithms of system become more and more sophisticated and complicated. The verification phase of knowledge base is an important part for developing reliable expert systems, especially in nuclear industry. Although several strategies or tools have been developed to perform potential error checking, they often neglect the reliability of verification methods. Because a Petri net provides a uniform mathematical formalization of knowledge base, it has been employed for knowledge base verification. In this work, we devise and suggest an automated tool, called COKEP(Checker of Knowledge base using Extended Petri net), for detecting incorrectness, inconsistency, and incompleteness in a knowledge base. The scope of the verification problem is expended to chained errors, unlike previous studies that assume error incidence to be limited to rule pairs only. In addition, we consider certainty factor in checking, because most of knowledge bases have certainly factors. 8 refs,. 2 figs,. 4 tabs. (author)

Verification of implementation of the radiological safety standards through the regulatory inspections

International Nuclear Information System (INIS)

Perez Gonzalez, Francisco; Fornet Rodriguez, Ofelia M.

2008-01-01

Full text: As an element of the updating process of the legal framework on radiological safety in Cuba, a new rule was put into force; the Radiological Basic Safety Standards (RBSS) in January 2002. Five years after the application of these new safety requirements, it was considered appropriate to assess the effectiveness of its implementation. Therefore, in this work the authors analysed the outcomes of the regulatory inspections conducted in this period upon medical and industrial practices in a sample of facilities representative of those with the highest radiological risks in the territory under supervision of a Territorial Delegation of the Nuclear Regulatory Authority. For better understanding of this presentation, a summary explanation of the structure of the rule is given in its introduction. The work was to identify for each deficiency, or finding, or counter-measure; out of the relevant inspections; the corresponding requirement/Article of the RBSS that shows difficulties in implementation. For each installation an analysis is made with regard to the relevant articles difficult to implement. Finally, the appraisal is shown separately for the medical practice, and for the industrial practice, and also in general for the whole sample of installations under review. The study showed that the implementation of the Standards has been satisfactory and uniform in the practices under review. So far it seems that there have not been major difficulties with the implementation of the Titles; III On Intervention, IV Dose Limits, as well as with the Especial, Final, and Transitory Dispositions. On the other hand, it is shown there is a need for continued work only with regard to the implementation of the requirements in Section IV Verification of Safety and in Section V On the responsibilities with regard to occupational exposure in Chapter III Title I, and correspondingly in Chapter II Occupational Exposure in Title II. It is recommended to conduct this kind of

Formal Verification Method for Configuration of Integrated Modular Avionics System Using MARTE

Directory of Open Access Journals (Sweden)

Lisong Wang

2018-01-01

Full Text Available The configuration information of Integrated Modular Avionics (IMA system includes almost all details of whole system architecture, which is used to configure the hardware interfaces, operating system, and interactions among applications to make an IMA system work correctly and reliably. It is very important to ensure the correctness and integrity of the configuration in the IMA system design phase. In this paper, we focus on modelling and verification of configuration information of IMA/ARINC653 system based on MARTE (Modelling and Analysis for Real-time and Embedded Systems. Firstly, we define semantic mapping from key concepts of configuration (such as modules, partitions, memory, process, and communications to components of MARTE element and propose a method for model transformation between XML-formatted configuration information and MARTE models. Then we present a formal verification framework for ARINC653 system configuration based on theorem proof techniques, including construction of corresponding REAL theorems according to the semantics of those key components of configuration information and formal verification of theorems for the properties of IMA, such as time constraints, spatial isolation, and health monitoring. After that, a special issue of schedulability analysis of ARINC653 system is studied. We design a hierarchical scheduling strategy with consideration of characters of the ARINC653 system, and a scheduling analyzer MAST-2 is used to implement hierarchical schedule analysis. Lastly, we design a prototype tool, called Configuration Checker for ARINC653 (CC653, and two case studies show that the methods proposed in this paper are feasible and efficient.

SFR Safety Considerations

International Nuclear Information System (INIS)

Glatz, Jean-Paul

2012-01-01

Objectives of the Safety and Operation Project: • analysis and experiments that support approaches and assess performance of specific safety features, • development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and • valorisation of reactor operation, from experience and testing in operating SFR plants

Technical Note: Range verification system using edge detection method for a scintillator and a CCD camera system

Energy Technology Data Exchange (ETDEWEB)

Saotome, Naoya, E-mail: naosao@nirs.go.jp; Furukawa, Takuji; Hara, Yousuke; Mizushima, Kota; Tansho, Ryohei; Saraya, Yuichi; Shirai, Toshiyuki; Noda, Koji [Department of Research Center for Charged Particle Therapy, National Institute of Radiological Sciences, 4-9-1 Anagawa, Inage-ku, Chiba 263-8555 (Japan)

2016-04-15

Purpose: Three-dimensional irradiation with a scanned carbon-ion beam has been performed from 2011 at the authors’ facility. The authors have developed the rotating-gantry equipped with the scanning irradiation system. The number of combinations of beam properties to measure for the commissioning is more than 7200, i.e., 201 energy steps, 3 intensities, and 12 gantry angles. To compress the commissioning time, quick and simple range verification system is required. In this work, the authors develop a quick range verification system using scintillator and charge-coupled device (CCD) camera and estimate the accuracy of the range verification. Methods: A cylindrical plastic scintillator block and a CCD camera were installed on the black box. The optical spatial resolution of the system is 0.2 mm/pixel. The camera control system was connected and communicates with the measurement system that is part of the scanning system. The range was determined by image processing. Reference range for each energy beam was determined by a difference of Gaussian (DOG) method and the 80% of distal dose of the depth-dose distribution that were measured by a large parallel-plate ionization chamber. The authors compared a threshold method and a DOG method. Results: The authors found that the edge detection method (i.e., the DOG method) is best for the range detection. The accuracy of range detection using this system is within 0.2 mm, and the reproducibility of the same energy measurement is within 0.1 mm without setup error. Conclusions: The results of this study demonstrate that the authors’ range check system is capable of quick and easy range verification with sufficient accuracy.

Technical Note: Range verification system using edge detection method for a scintillator and a CCD camera system

International Nuclear Information System (INIS)

Saotome, Naoya; Furukawa, Takuji; Hara, Yousuke; Mizushima, Kota; Tansho, Ryohei; Saraya, Yuichi; Shirai, Toshiyuki; Noda, Koji

2016-01-01

Purpose: Three-dimensional irradiation with a scanned carbon-ion beam has been performed from 2011 at the authors’ facility. The authors have developed the rotating-gantry equipped with the scanning irradiation system. The number of combinations of beam properties to measure for the commissioning is more than 7200, i.e., 201 energy steps, 3 intensities, and 12 gantry angles. To compress the commissioning time, quick and simple range verification system is required. In this work, the authors develop a quick range verification system using scintillator and charge-coupled device (CCD) camera and estimate the accuracy of the range verification. Methods: A cylindrical plastic scintillator block and a CCD camera were installed on the black box. The optical spatial resolution of the system is 0.2 mm/pixel. The camera control system was connected and communicates with the measurement system that is part of the scanning system. The range was determined by image processing. Reference range for each energy beam was determined by a difference of Gaussian (DOG) method and the 80% of distal dose of the depth-dose distribution that were measured by a large parallel-plate ionization chamber. The authors compared a threshold method and a DOG method. Results: The authors found that the edge detection method (i.e., the DOG method) is best for the range detection. The accuracy of range detection using this system is within 0.2 mm, and the reproducibility of the same energy measurement is within 0.1 mm without setup error. Conclusions: The results of this study demonstrate that the authors’ range check system is capable of quick and easy range verification with sufficient accuracy.

The Integrated Safety Management System Verification Enhancement Review of the Plutonium Finishing Plant (PFP)

International Nuclear Information System (INIS)

BRIGGS, C.R.

2000-01-01

The primary purpose of the verification enhancement review was for the DOE Richland Operations Office (RL) to verify contractor readiness for the independent DOE Integrated Safety Management System Verification (ISMSV) on the Plutonium Finishing Plant (PFP). Secondary objectives included: (1) to reinforce the engagement of management and to gauge management commitment and accountability; (2) to evaluate the ''value added'' benefit of direct public involvement; (3) to evaluate the ''value added'' benefit of direct worker involvement; (4) to evaluate the ''value added'' benefit of the panel-to-panel review approach; and, (5) to evaluate the utility of the review's methodology/adaptability to periodic assessments of ISM status. The review was conducted on December 6-8, 1999, and involved the conduct of two-hour interviews with five separate panels of individuals with various management and operations responsibilities related to PFP. A semi-structured interview process was employed by a team of five ''reviewers'' who directed open-ended questions to the panels which focused on: (1) evidence of management commitment, accountability, and involvement; and, (2) consideration and demonstration of stakeholder (including worker) information and involvement opportunities. The purpose of a panel-to-panel dialogue approach was to better spotlight: (1) areas of mutual reinforcement and alignment that could serve as good examples of the management commitment and accountability aspects of ISMS implementation, and, (2) areas of potential discrepancy that could provide opportunities for improvement. In summary, the Review Team found major strengths to include: (1) the use of multi-disciplinary project work teams to plan and do work; (2) the availability and broad usage of multiple tools to help with planning and integrating work; (3) senior management presence and accessibility; (4) the institutionalization of worker involvement; (5) encouragement of self-reporting and self

Verification of FPGA-based NPP I and C systems. General approach and techniques

International Nuclear Information System (INIS)

Andrashov, Anton; Kharchenko, Vyacheslav; Sklyar, Volodymir; Reva, Lubov; Siora, Alexander

2011-01-01

This paper presents a general approach and techniques for design and verification of Field Programmable Gates Arrays (FPGA)-based Instrumentation and Control (I and C) systems for Nuclear Power Plants (NPP). Appropriate regulatory documents used for I and C systems design, development, verification and validation (V and V) are discussed considering the latest international standards and guidelines. Typical development and V and V processes of FPGA electronic design for FPGA-based NPP I and C systems are presented. Some safety-related features of implementation process are discussed. Corresponding development artifacts, related to design and implementation activities are outlined. An approach to test-based verification of FPGA electronic design algorithms, used in FPGA-based reactor trip systems is proposed. The results of application of test-based techniques for assessment of FPGA electronic design algorithms for reactor trip system (RTS) produced by Research and Production Corporation (RPC) 'Radiy' are presented. Some principles of invariant-oriented verification for FPGA-based safety-critical systems are outlined. (author)

Tools and Methods for RTCP-Nets Modeling and Verification

Directory of Open Access Journals (Sweden)

Szpyrka Marcin

2016-09-01

Full Text Available RTCP-nets are high level Petri nets similar to timed colored Petri nets, but with different time model and some structural restrictions. The paper deals with practical aspects of using RTCP-nets for modeling and verification of real-time systems. It contains a survey of software tools developed to support RTCP-nets. Verification of RTCP-nets is based on coverability graphs which represent the set of reachable states in the form of directed graph. Two approaches to verification of RTCP-nets are considered in the paper. The former one is oriented towards states and is based on translation of a coverability graph into nuXmv (NuSMV finite state model. The later approach is oriented towards transitions and uses the CADP toolkit to check whether requirements given as μ-calculus formulae hold for a given coverability graph. All presented concepts are discussed using illustrative examples

SU-F-T-440: The Feasibility Research of Checking Cervical Cancer IMRT Pre- Treatment Dose Verification by Automated Treatment Planning Verification System

Energy Technology Data Exchange (ETDEWEB)

Liu, X; Yin, Y; Lin, X [Shandong Cancer Hospital and Institute, China, Jinan, Shandong (China)

2016-06-15

Purpose: To assess the preliminary feasibility of automated treatment planning verification system in cervical cancer IMRT pre-treatment dose verification. Methods: The study selected randomly clinical IMRT treatment planning data for twenty patients with cervical cancer, all IMRT plans were divided into 7 fields to meet the dosimetric goals using a commercial treatment planning system(PianncleVersion 9.2and the EclipseVersion 13.5). The plans were exported to the Mobius 3D (M3D)server percentage differences of volume of a region of interest (ROI) and dose calculation of target region and organ at risk were evaluated, in order to validate the accuracy automated treatment planning verification system. Results: The difference of volume for Pinnacle to M3D was less than results for Eclipse to M3D in ROI, the biggest difference was 0.22± 0.69%, 3.5±1.89% for Pinnacle and Eclipse respectively. M3D showed slightly better agreement in dose of target and organ at risk compared with TPS. But after recalculating plans by M3D, dose difference for Pinnacle was less than Eclipse on average, results were within 3%. Conclusion: The method of utilizing the automated treatment planning system to validate the accuracy of plans is convenientbut the scope of differences still need more clinical patient cases to determine. At present, it should be used as a secondary check tool to improve safety in the clinical treatment planning.

Formal Verification of Digital Protection Logic and Automatic Testing Software

Energy Technology Data Exchange (ETDEWEB)

Cha, S. D.; Ha, J. S.; Seo, J. S. [KAIST, Daejeon (Korea, Republic of)

2008-06-15

- Technical aspect {center_dot} It is intended that digital I and C software have safety and reliability. Project results help the software to acquire license. Software verification technique, which results in this project, can be to use for digital NPP(Nuclear power plant) in the future. {center_dot} This research introduces many meaningful results of verification on digital protection logic and suggests I and C software testing strategy. These results apply to verify nuclear fusion device, accelerator, nuclear waste management and nuclear medical device that require dependable software and high-reliable controller. Moreover, These can be used for military, medical or aerospace-related software. - Economical and industrial aspect {center_dot} Since safety of digital I and C software is highly import, It is essential for the software to be verified. But verification and licence acquisition related to digital I and C software face high cost. This project gives economic profit to domestic economy by using introduced verification and testing technique instead of foreign technique. {center_dot} The operation rate of NPP will rise, when NPP safety critical software is verified with intellectual V and V tool. It is expected that these software substitute safety-critical software that wholly depend on foreign. Consequently, the result of this project has high commercial value and the recognition of the software development works will be able to be spread to the industrial circles. - Social and cultural aspect People expect that nuclear power generation contributes to relieving environmental problems because that does not emit more harmful air pollution source than other power generations. To give more trust and expectation about nuclear power generation to our society, we should make people to believe that NPP is highly safe system. In that point of view, we can present high-reliable I and C proofed by intellectual V and V technique as evidence

Development of safety enhancement technology of containment building

International Nuclear Information System (INIS)

Seo, Jeong Moon; Choun, Y. S.; Choi, I. K.

2002-04-01

This study consists of four research areas, (1) Seismic safety assessment, (2) Aging assessment of a containment building, (3) Prediction of long-term behavior and analysis of a containment building, (4) Performance verification of a containment building. In the seismic safety assessment area, responses of a containment building were monitored and the analysis method was verified. Also performed are the identification of earthquake characteristics and improvement of the seismic fragility analysis method. In the area of aging assessment of a containment building, we developed aging management code SLMS and database. Aging tests were performed for containment building materials and aging models were developed. Techniques for investigation, detection, and evaluation of aging were developed. In the area of prediction of long-term behavior and analysis of a containment building, we developed a non-linear structural analysis code NUCAS and material models. In the area of performance verification of a containment building, we analyzed the crack behavior of a containment wall and the behavior of the containment under internal pressure. We also improved the ISI methods for prestressed containment

Core power capability verification for PWR NPP

International Nuclear Information System (INIS)

Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

2002-01-01

The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

Theoretical interpretations and experimental verifications of a radioelectric resonance method for measuring the electronic density and collision frequency in a discharge plasma in gases; Interpretations theoriques et verifications experimentales d'une methode de resonance radioelectrique pour la mesure de la densite d'une decharge dans les gaz

Energy Technology Data Exchange (ETDEWEB)

Nguyen Trong, Khoi [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

Theoretical discussions and experimental verifications of one radioelectric resonance method for measuring plasma electronic density and collision frequency. (author) [French] Discussions theoriques et verifications experimentales sur une methode de resonance radioelectrique pour la mesure de la densite electronique et de la frequence de collision d'un plasma d'une decharge dans le gaz. (auteur)

Nuclear Data Verification and Standardization

Energy Technology Data Exchange (ETDEWEB)

Karam, Lisa R.; Arif, Muhammad; Thompson, Alan K.

2011-10-01

The objective of this interagency program is to provide accurate neutron interaction verification and standardization data for the U.S. Department of Energy Division of Nuclear Physics programs which include astrophysics, radioactive beam studies, and heavy-ion reactions. The measurements made in this program are also useful to other programs that indirectly use the unique properties of the neutron for diagnostic and analytical purposes. These include homeland security, personnel health and safety, nuclear waste disposal, treaty verification, national defense, and nuclear based energy production. The work includes the verification of reference standard cross sections and related neutron data employing the unique facilities and capabilities at NIST and other laboratories as required; leadership and participation in international intercomparisons and collaborations; and the preservation of standard reference deposits. An essential element of the program is critical evaluation of neutron interaction data standards including international coordinations. Data testing of critical data for important applications is included. The program is jointly supported by the Department of Energy and the National Institute of Standards and Technology.

Whole-core thermal-hydraulic transient code development and verification for LMFBR analysis

International Nuclear Information System (INIS)

Spencer, D.R.

1979-04-01

Predicted performance during both steady state and transient reactor operation determines the steady state operating limits on LMFBRs. Unnecessary conservatism in performance predictions will not contribute to safety, but will restrict the reactor to more conservative, less economical steady state operation. The most general method for reducing analytical conservatism in LMFBR's without compromising safety is to develop, validate and apply more sophisticated computer models to the limiting performance analyses. The purpose of the on-going Natural Circulation Verification Program (NCVP) is to develop and validate computer codes to analyze natural circulation transients in LMFBRs, and thus, replace unnecessary analytical conservatism with demonstrated calculational capability

Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues. Supplement 4

International Nuclear Information System (INIS)

1994-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. Supplement 3 gives status as of September 30, 1993. This annual report, Supplement 4, presents updated information as of September 30, 1994. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

Methodology and tools for independent verification and validation of computerized I and C systems important to safety

International Nuclear Information System (INIS)

Lindner, A.; Miedl, H.

1998-01-01

Modular software based I and C systems are state-of-the-art in industrial automation. For I and C systems important to safety in nuclear power plants, software based systems are also more and more applied. According to existing national and international guidelines and standards, the assessment of these systems calls for appropriate test methods and tools. By use of tools quality of the assessment process should be improved and expense should be limited. The paper outlines the structure of the independent verification and validation (V and V) process of the Teleperm XS system and the lessons learnt from this process. Furthermore, tools are discussed used for V and V of the Teleperm XS software. The recently developed tool VALIDATOR, dedicated to V and V of the plant specific I and C functions is described in more detail. We consider V and V of the basic software components and the system software to be required only once, but the C source codes of the plant specific functional diagrams have to be checked for each application separately. The VALIDATOR is designed to perform this task. It gives evidence of compliance of the automatically generated C source codes with the graphical design of the functional diagrams in reasonable time and with acceptable costs. The working method, performance and results of the VALIDATOR are shown by means of an actual example. (author)

Towards Verification of Operational Procedures Using Auto-Generated Diagnostic Trees

Science.gov (United States)

Kurtoglu, Tolga; Lutz, Robyn; Patterson-Hine, Ann

2009-01-01

The design, development, and operation of complex space, lunar and planetary exploration systems require the development of general procedures that describe a detailed set of instructions capturing how mission tasks are performed. For both crewed and uncrewed NASA systems, mission safety and the accomplishment of the scientific mission objectives are highly dependent on the correctness of procedures. In this paper, we describe how to use the auto-generated diagnostic trees from existing diagnostic models to improve the verification of standard operating procedures. Specifically, we introduce a systematic method, namely the Diagnostic Tree for Verification (DTV), developed with the goal of leveraging the information contained within auto-generated diagnostic trees in order to check the correctness of procedures, to streamline the procedures in terms of reducing the number of steps or use of resources in them, and to propose alternative procedural steps adaptive to changing operational conditions. The application of the DTV method to a spacecraft electrical power system shows the feasibility of the approach and its range of capabilities

Standard practice for verification and classification of extensometer systems

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for the verification and classification of extensometer systems, but it is not intended to be a complete purchase specification. The practice is applicable only to instruments that indicate or record values that are proportional to changes in length corresponding to either tensile or compressive strain. Extensometer systems are classified on the basis of the magnitude of their errors. 1.2 Because strain is a dimensionless quantity, this document can be used for extensometers based on either SI or US customary units of displacement. Note 1—Bonded resistance strain gauges directly bonded to a specimen cannot be calibrated or verified with the apparatus described in this practice for the verification of extensometers having definite gauge points. (See procedures as described in Test Methods E251.) 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish app...

A Correctness Verification Technique for Commercial FPGA Synthesis Tools

International Nuclear Information System (INIS)

Kim, Eui Sub; Yoo, Jun Beom; Choi, Jong Gyun; Kim, Jang Yeol; Lee, Jang Soo

2014-01-01

Once the FPGA (Filed-Programmable Gate Array) designers designs Verilog programs, the commercial synthesis tools automatically translate the Verilog programs into EDIF programs so that the designers can have largely focused on HDL designs for correctness of functionality. Nuclear regulation authorities, however, require more considerate demonstration of the correctness and safety of mechanical synthesis processes of FPGA synthesis tools, even if the FPGA industry have acknowledged them empirically as correct and safe processes and tools. In order to assure of the safety, the industry standards for the safety of electronic/electrical devices, such as IEC 61508 and IEC 60880, recommend using the formal verification technique. There are several formal verification tools (i.e., 'FormalPro' 'Conformal' 'Formality' and so on) to verify the correctness of translation from Verilog into EDIF programs, but it is too expensive to use and hard to apply them to the works of 3rd-party developers. This paper proposes a formal verification technique which can contribute to the correctness demonstration in part. It formally checks the behavioral equivalence between Verilog and subsequently synthesized Net list with the VIS verification system. A Net list is an intermediate output of FPGA synthesis process, and EDIF is used as a standard format of Net lists. If the formal verification succeeds, then we can assure that the synthesis process from Verilog into Net list worked correctly at least for the Verilog used. In order to support the formal verification, we developed the mechanical translator 'EDIFtoBLIFMV,' which translates EDIF into BLIF-MV as an input front-end of VIS system, while preserving their behavior equivalence.. We performed the case study with an example of a preliminary version of RPS in a Korean nuclear power plant in order to provide the efficiency of the proposed formal verification technique and implemented translator. It

A Correctness Verification Technique for Commercial FPGA Synthesis Tools

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Sub; Yoo, Jun Beom [Konkuk University, Seoul (Korea, Republic of); Choi, Jong Gyun; Kim, Jang Yeol; Lee, Jang Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Once the FPGA (Filed-Programmable Gate Array) designers designs Verilog programs, the commercial synthesis tools automatically translate the Verilog programs into EDIF programs so that the designers can have largely focused on HDL designs for correctness of functionality. Nuclear regulation authorities, however, require more considerate demonstration of the correctness and safety of mechanical synthesis processes of FPGA synthesis tools, even if the FPGA industry have acknowledged them empirically as correct and safe processes and tools. In order to assure of the safety, the industry standards for the safety of electronic/electrical devices, such as IEC 61508 and IEC 60880, recommend using the formal verification technique. There are several formal verification tools (i.e., 'FormalPro' 'Conformal' 'Formality' and so on) to verify the correctness of translation from Verilog into EDIF programs, but it is too expensive to use and hard to apply them to the works of 3rd-party developers. This paper proposes a formal verification technique which can contribute to the correctness demonstration in part. It formally checks the behavioral equivalence between Verilog and subsequently synthesized Net list with the VIS verification system. A Net list is an intermediate output of FPGA synthesis process, and EDIF is used as a standard format of Net lists. If the formal verification succeeds, then we can assure that the synthesis process from Verilog into Net list worked correctly at least for the Verilog used. In order to support the formal verification, we developed the mechanical translator 'EDIFtoBLIFMV,' which translates EDIF into BLIF-MV as an input front-end of VIS system, while preserving their behavior equivalence.. We performed the case study with an example of a preliminary version of RPS in a Korean nuclear power plant in order to provide the efficiency of the proposed formal verification technique and implemented translator. It

Safety evaluations required in the safety regulations for Monju and the validity confirmation of safety evaluation methods

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purposes of this study are to perform the safety evaluations of the fast breeder reactor 'Monju' and to confirm the validity of the safety evaluation methods. In JFY 2012, the following results were obtained. As for the development of safety evaluation methods needed in the safety examination achieved for the reactor establishment permission, development of the analysis codes, such as a core damage analysis code, were carried out according to the plan. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

Formal specification and verification of interactive systems with plasticity: Applications to nuclear-plant supervision

International Nuclear Information System (INIS)

Oliveira, Raquel Araujo de

2015-01-01

The advent of ubiquitous computing and the increasing variety of platforms and devices change user expectations in terms of user interfaces. Systems should be able to adapt themselves to their context of use, i.e., the platform (e.g. a PC or a tablet), the users who interact with the system (e.g. administrators or regular users), and the environment in which the system executes (e.g. a dark room or outdoor). The capacity of a UI to withstand variations in its context of use while preserving usability is called plasticity. Plasticity provides users with different versions of a UI. Although it enhances UI capabilities, plasticity adds complexity to the development of user interfaces: the consistency between multiple versions of a given UI should be ensured. Given the large number of possible versions of a UI, it is time-consuming and error prone to check these requirements by hand. Some automation must be provided to verify plasticity.This complexity is further increased when it comes to UIs of safety-critical systems. Safety-critical systems are systems in which a failure has severe consequences. The complexity of such systems is reflected in the UIs, which are now expected not only to provide correct, intuitive, non-ambiguous and adaptable means for users to accomplish a goal, but also to cope with safety requirements aiming to make sure that systems are reasonably safe before they enter the market. Several techniques to ensure quality of systems in general exist, which can also be used to safety-critical systems. Formal verification provides a rigorous way to perform verification, which is suitable for safety-critical systems. Our contribution is an approach to verify safety-critical interactive systems provided with plastic UIs using formal methods. Using a powerful tool-support, our approach permits:-The verification of sets of properties over a model of the system. Using model checking, our approach permits the verification of properties over the system formal

Quantifying system safety: A comparison of the SBOAT & Safety Barrier Manager tools

DEFF Research Database (Denmark)

Hansen, Zaza Nadja Lee; Duijm, Nijs Jan; Markert, Frank

2015-01-01

This paper presents two software tools for analyzing safety risks, SBOAT (Stochastic BPMN Optimisation and Analysis Tool) and SBM (SafetyBarrierManagerr). SBOAT employs principles from stochastic model checking to allow for the quantitative verification of workflows. SBM supports the creation...

KNGR core proection calculator, software, verification and validation plan

International Nuclear Information System (INIS)

Kim, Jang Yeol; Park, Jong Kyun; Lee, Ki Young; Lee, Jang Soo; Cheon, Se Woo

2001-05-01

This document describes the Software Verification and Validation Plan(SVVP) Guidance to be used in reviewing the Software Program Manual(SPM) in Korean Next Generation Reactor(KNGR) projects. This document is intended for a verifier or reviewer who is involved with performing of software verification and validation task activity in KNGR projects. This document includeds the basic philosophy, performing V and V effort, software testing techniques, criteria of review and audit on the safety software V and V activity. Major review topics on safety software addresses three kinds of characteristics based on Standard Review Plan(SRP) Chapter 7, Branch Technical Position(BTP)-14 : management characteristics, implementation characteristics and resources characteristics when reviewing on SVVP. Based on major topics of this document, we have produced the evaluation items list such as checklist in Appendix A

FMEF Electrical single line diagram and panel schedule verification process

International Nuclear Information System (INIS)

Fong, S.K.

1998-01-01

Since the FMEF did not have a mission, a formal drawing verification program was not developed, however, a verification process on essential electrical single line drawings and panel schedules was established to benefit the operations lock and tag program and to enhance the electrical safety culture of the facility. The purpose of this document is to provide a basis by which future landlords and cognizant personnel can understand the degree of verification performed on the electrical single lines and panel schedules. It is the intent that this document be revised or replaced by a more formal requirements document if a mission is identified for the FMEF

Case Study: Test Results of a Tool and Method for In-Flight, Adaptive Control System Verification on a NASA F-15 Flight Research Aircraft

Science.gov (United States)

Jacklin, Stephen A.; Schumann, Johann; Guenther, Kurt; Bosworth, John

2006-01-01

Adaptive control technologies that incorporate learning algorithms have been proposed to enable autonomous flight control and to maintain vehicle performance in the face of unknown, changing, or poorly defined operating environments [1-2]. At the present time, however, it is unknown how adaptive algorithms can be routinely verified, validated, and certified for use in safety-critical applications. Rigorous methods for adaptive software verification end validation must be developed to ensure that. the control software functions as required and is highly safe and reliable. A large gap appears to exist between the point at which control system designers feel the verification process is complete, and when FAA certification officials agree it is complete. Certification of adaptive flight control software verification is complicated by the use of learning algorithms (e.g., neural networks) and degrees of system non-determinism. Of course, analytical efforts must be made in the verification process to place guarantees on learning algorithm stability, rate of convergence, and convergence accuracy. However, to satisfy FAA certification requirements, it must be demonstrated that the adaptive flight control system is also able to fail and still allow the aircraft to be flown safely or to land, while at the same time providing a means of crew notification of the (impending) failure. It was for this purpose that the NASA Ames Confidence Tool was developed [3]. This paper presents the Confidence Tool as a means of providing in-flight software assurance monitoring of an adaptive flight control system. The paper will present the data obtained from flight testing the tool on a specially modified F-15 aircraft designed to simulate loss of flight control faces.

Explosion overpressure test series: General-Purpose Heat Source development: Safety Verification Test program

International Nuclear Information System (INIS)

Cull, T.A.; George, T.G.; Pavone, D.

1986-09-01

The General-Purpose Heat Source (GPHS) is a modular, radioisotope heat source that will be used in radioisotope thermoelectric generators (RTGs) to supply electric power for space missions. The first two uses will be the NASA Galileo and the ESA Ulysses missions. The RTG for these missions will contain 18 GPHS modules, each of which contains four 238 PuO 2 -fueled clads and generates 250 W/sub (t)/. A series of Safety Verification Tests (SVTs) was conducted to assess the ability of the GPHS modules to contain the plutonia in accident environments. Because a launch pad or postlaunch explosion of the Space Transportation System vehicle (space shuttle) is a conceivable accident, the SVT plan included a series of tests that simulated the overpressure exposure the RTG and GPHS modules could experience in such an event. Results of these tests, in which we used depleted UO 2 as a fuel simulant, suggest that exposure to overpressures as high as 15.2 MPa (2200 psi), without subsequent impact, does not result in a release of fuel

Concepts for inventory verification in critical facilities

International Nuclear Information System (INIS)

Cobb, D.D.; Sapir, J.L.; Kern, E.A.; Dietz, R.J.

1978-12-01

Materials measurement and inventory verification concepts for safeguarding large critical facilities are presented. Inspection strategies and methods for applying international safeguards to such facilities are proposed. The conceptual approach to routine inventory verification includes frequent visits to the facility by one inspector, and the use of seals and nondestructive assay (NDA) measurements to verify the portion of the inventory maintained in vault storage. Periodic verification of the reactor inventory is accomplished by sampling and NDA measurement of in-core fuel elements combined with measurements of integral reactivity and related reactor parameters that are sensitive to the total fissile inventory. A combination of statistical sampling and NDA verification with measurements of reactor parameters is more effective than either technique used by itself. Special procedures for assessment and verification for abnormal safeguards conditions are also considered. When the inspection strategies and inventory verification methods are combined with strict containment and surveillance methods, they provide a high degree of assurance that any clandestine attempt to divert a significant quantity of fissile material from a critical facility inventory will be detected. Field testing of specific hardware systems and procedures to determine their sensitivity, reliability, and operational acceptability is recommended. 50 figures, 21 tables

Formal Modeling and Verification of Interlocking Systems Featuring Sequential Release

DEFF Research Database (Denmark)

Vu, Linh Hong; Haxthausen, Anne Elisabeth; Peleska, Jan

2015-01-01

In this paper, we present a method and an associated tool suite for formal verification of the new ETCS level 2 based Danish railway interlocking systems. We have made a generic and reconfigurable model of the system behavior and generic high-level safety properties. This model accommodates seque...... SMT based bounded model checking (BMC) and inductive reasoning, we are able to verify the properties for model instances corresponding to railway networks of industrial size. Experiments also show that BMC is efficient for finding bugs in the railway interlocking designs....

Formal Modeling and Verification of Interlocking Systems Featuring Sequential Release

DEFF Research Database (Denmark)

Vu, Linh Hong; Haxthausen, Anne Elisabeth; Peleska, Jan

2014-01-01

In this paper, we present a method and an associated tool suite for formal verification of the new ETCS level 2 based Danish railway interlocking systems. We have made a generic and reconfigurable model of the system behavior and generic high-level safety properties. This model accommodates seque...... SMT based bounded model checking (BMC) and inductive reasoning, we are able to verify the properties for model instances corresponding to railway networks of industrial size. Experiments also show that BMC is efficient for finding bugs in the railway interlocking designs....

Safety verification for the ECCS driven by the electrically 4 trains during LBLOCA reflood phase using ATLAS

International Nuclear Information System (INIS)

Park, Yusun; Park, Hyun-sik; Kang, Kyoung-ho; Choi, Nam-hyun; Min, Kyoung-ho; Choi, Ki-yong

2014-01-01

Highlights: • Safety improvement by adopting 4 train emergency core cooling system was validated experimentally. • General thermal hydraulic behaviors of the system during LBLOCA reflood phase were successfully demonstrated. • Key parameters such as the liquid levels, the PCTs, the quenching time, and the ECC bypass ratios were investigated. • Asymmetric effects of the different combination of safety injection were negligible during the reflood period. - Abstract: The APR1400 is equipped with four safety injection pumps driven by two emergency diesel generators. However, the design has been changed so that the four safety injection pumps are driven by 4 emergency diesel generators during the design certification process from the U.S. NRC. Thus, 4 safety injection pumps (SIPs) are completely independent electrically and mechanically and three safety injection pumps are available in a single failure condition. This design change could have a certain effects on the thermal-hydraulic phenomenon occurring in the downcomer region during the late reflood phase of a large break loss of coolant accident (LBLOCA). Thus, in this study, a verification experiment for the reflood phase of a LBLOCA was performed to evaluate the core cooling performance of the 4 train emergency core cooling system (ECCS) with an assumption of a single failure. And the different combinations of three SIPs positions were tested to investigate the asymmetric effects on the reactor core cooling performance. The overall experimental results revealed the typical thermal–hydraulic trends expected to occur during the reflood phase of a large-break LOCA scenario for the APR1400. Experiment with the injection of three SIPs showed a faster core quenching time and lower bypass ratio than that of the case in which two SIPs were injected. The RPV wall temperature distributions showed the similar trend in spite of the different SIP combinations

Mechanical verification of concurrency control and recovery protocols

NARCIS (Netherlands)

Chkliaev, D.

2001-01-01

The thesis concerns the formal specification and mechanized verification of concurrency control and recovery protocols for distributed databases. Such protocols are needed for many modern application such as banking and are often used in safety-critical applications. Therefore it is very important

Validation of the implementation of IMRT with three dosimetric methods of independent verification

International Nuclear Information System (INIS)

Tortosa Oliver, R. A.; Chinillach ferrando, N.; Alonso Arrizabalaga, S.; Campayo Esteban, J. M.; Morales Marco, J. C.; Soler Catalan, P.; Andreu Martinez, F. J.

2013-01-01

The TG119 is a simple and clear framework to verify the implementation of IMRT technique in a radiotherapy service. Verifications of this document recommended tests conducted with the three dosimetric methods listed above, allow to affirm that our Center is within the margins of tolerance considered suitable in the TG119 for the clinical implementation of IMRT. (Author)

Verification of Bioanalytical Method for Quantification of Exogenous Insulin (Insulin Aspart) by the Analyser Advia Centaur® XP.

Science.gov (United States)

Mihailov, Rossen; Stoeva, Dilyana; Pencheva, Blagovesta; Pentchev, Eugeni

2018-03-01

In a number of cases the monitoring of patients with type I diabetes mellitus requires measurement of the exogenous insulin levels. For the purpose of a clinical investigation of the efficacy of a medical device for application of exogenous insulin aspart, a verification of the method for measurement of this synthetic analogue of the hormone was needed. The information in the available medical literature for the measurement of the different exogenous insulin analogs is insufficient. Thus, verification was required to be in compliance with the active standards in Republic of Bulgaria. A manufactured method developed for ADVIA Centaur XP Immunoassay, Siemens Healthcare, was used which we verified using standard solutions and a patient serum pool by adding the appropriate quantity exogenous insulin aspart. The method was verified in accordance with the bioanalytical method verification criteria and regulatory requirements for using a standard method: CLIA chemiluminescence immunoassay ADVIA Centaur® XP. The following parameters are determined and monitored: intra-day precision and accuracy, inter-day precision and accuracy, limit of detection and lower limit of quantification, linearity, analytical recovery. The routine application of the method for measurement of immunoreactive insulin using the analyzer ADVIA Centaur® XP is directed to the measurement of endogenous insulin. The method is applicable for measuring different types of exogenous insulin, including insulin aspart.

Specification and Automated Verification of Real-Time Behaviour

DEFF Research Database (Denmark)

Kristensen, C.H.; Andersen, J.H.; Skou, A.

1995-01-01

In this paper we sketch a method for specification and automatic verification of real-time software properties.......In this paper we sketch a method for specification and automatic verification of real-time software properties....

Specification and Automated Verification of Real-Time Behaviour

DEFF Research Database (Denmark)

Andersen, J.H.; Kristensen, C.H.; Skou, A.

1996-01-01

In this paper we sketch a method for specification and automatic verification of real-time software properties.......In this paper we sketch a method for specification and automatic verification of real-time software properties....

SU-E-I-24: Method for CT Automatic Exposure Control Verification

Energy Technology Data Exchange (ETDEWEB)

Gracia, M; Olasolo, J; Martin, M; Bragado, L; Gallardo, N; Miquelez, S; Maneru, F; Lozares, S; Pellejero, S; Rubio, A [Complejo Hospitalario de Navarra, Pamplona, Navarra (Spain)

2015-06-15

Purpose: Design of a phantom and a simple method for the automatic exposure control (AEC) verification in CT. This verification is included in the computed tomography (CT) Spanish Quality Assurance Protocol. Methods: The phantom design is made from the head and the body phantom used for the CTDI measurement and PMMA plates (35×35 cm2) of 10 cm thickness. Thereby, three different thicknesses along the longitudinal axis are obtained which permit to evaluate the longitudinal AEC performance. Otherwise, the existent asymmetry in the PMMA layers helps to assess angular and 3D AEC operation.Recent acquisition in our hospital (August 2014) of Nomex electrometer (PTW), together with the 10 cm pencil ionization chamber, led to register dose rate as a function of time. Measurements with this chamber fixed at 0° and 90° on the gantry where made on five multidetector-CTs from principal manufacturers. Results: Individual analysis of measurements shows dose rate variation as a function of phantom thickness. The comparative analysis shows that dose rate is kept constant in the head and neck phantom while the PMMA phantom exhibits an abrupt variation between both results, being greater results at 90° as the thickness of the phantom is 3.5 times larger than in the perpendicular direction. Conclusion: Proposed method is simple, quick and reproducible. Results obtained let a qualitative evaluation of the AEC and they are consistent with the expected behavior. A line of future development is to quantitatively study the intensity modulation and parameters of image quality, and a possible comparative study between different manufacturers.

Nuclear disarmament verification

International Nuclear Information System (INIS)

DeVolpi, A.

1993-01-01

Arms control treaties, unilateral actions, and cooperative activities -- reflecting the defusing of East-West tensions -- are causing nuclear weapons to be disarmed and dismantled worldwide. In order to provide for future reductions and to build confidence in the permanency of this disarmament, verification procedures and technologies would play an important role. This paper outlines arms-control objectives, treaty organization, and actions that could be undertaken. For the purposes of this Workshop on Verification, nuclear disarmament has been divided into five topical subareas: Converting nuclear-weapons production complexes, Eliminating and monitoring nuclear-weapons delivery systems, Disabling and destroying nuclear warheads, Demilitarizing or non-military utilization of special nuclear materials, and Inhibiting nuclear arms in non-nuclear-weapons states. This paper concludes with an overview of potential methods for verification

Class 1E software verification and validation: Past, present, and future

Energy Technology Data Exchange (ETDEWEB)

Persons, W.L.; Lawrence, J.D.

1993-10-01

This paper discusses work in progress that addresses software verification and validation (V&V) as it takes place during the full software life cycle of safety-critical software. The paper begins with a brief overview of the task description and discussion of the historical evolution of software V&V. A new perspective is presented which shows the entire verification and validation process from the viewpoints of a software developer, product assurance engineer, independent V&V auditor, and government regulator. An account of the experience of the field test of the Verification Audit Plan and Report generated from the V&V Guidelines is presented along with sample checklists and lessons learned from the verification audit experience. Then, an approach to automating the V&V Guidelines is introduced. The paper concludes with a glossary and bibliography.

Verification of spectrophotometric method for nitrate analysis in water samples

Science.gov (United States)

Kurniawati, Puji; Gusrianti, Reny; Dwisiwi, Bledug Bernanti; Purbaningtias, Tri Esti; Wiyantoko, Bayu

2017-12-01

The aim of this research was to verify the spectrophotometric method to analyze nitrate in water samples using APHA 2012 Section 4500 NO3-B method. The verification parameters used were: linearity, method detection limit, level of quantitation, level of linearity, accuracy and precision. Linearity was obtained by using 0 to 50 mg/L nitrate standard solution and the correlation coefficient of standard calibration linear regression equation was 0.9981. The method detection limit (MDL) was defined as 0,1294 mg/L and limit of quantitation (LOQ) was 0,4117 mg/L. The result of a level of linearity (LOL) was 50 mg/L and nitrate concentration 10 to 50 mg/L was linear with a level of confidence was 99%. The accuracy was determined through recovery value was 109.1907%. The precision value was observed using % relative standard deviation (%RSD) from repeatability and its result was 1.0886%. The tested performance criteria showed that the methodology was verified under the laboratory conditions.

Office of River Protection Integrated Safety Management System Phase 1 Verification Corrective Action Plan

International Nuclear Information System (INIS)

CLARK, D.L.

1999-01-01

The purpose of this Corrective Action Plan is to demonstrate the OW planned and/or completed actions to implement ISMS as well as prepare for the RPP ISMS Phase II Verification scheduled for August, 1999. This Plan collates implied or explicit ORP actions identified in several key ISMS documents and aligns those actions and responsibilities perceived necessary to appropriately disposition all ISM Phase II preparation activities specific to the ORP. The objective will be to complete or disposition the corrective actions prior to the commencement of the ISMS Phase II Verification. Improvement products/tasks not slated for completion prior to the RPP Phase II verification will be incorporated as corrective actions into the Strategic System Execution Plan (SSEP) Gap Analysis. Many of the business and management systems that were reviewed in the ISMS Phase I verification are being modified to support the ORP transition and are being assessed through the SSEP. The actions and processes identified in the SSEP will support the development of the ORP and continued ISMS implementation as committed to be complete by end of FY-2000

Development of requirements tracking and verification technology for the NPP software

Energy Technology Data Exchange (ETDEWEB)

Jung, Chul Hwan; Kim, Jang Yeol; Lee, Jang Soo; Song, Soon Ja; Lee, Dong Young; Kwon, Kee Choon

1998-12-30

Searched and analyzed the technology of requirements engineering in the areas of aerospace and defense industry, medical industry and nuclear industry. Summarized the status of tools for the software design and requirements management. Analyzed the software design methodology for the safety software of NPP. Development of the design requirements for the requirements tracking and verification system. Development of the background technology to design the prototype tool for the requirements tracking and verification.

Development of requirements tracking and verification technology for the NPP software

International Nuclear Information System (INIS)

Jung, Chul Hwan; Kim, Jang Yeol; Lee, Jang Soo; Song, Soon Ja; Lee, Dong Young; Kwon, Kee Choon

1998-01-01

Searched and analyzed the technology of requirements engineering in the areas of aerospace and defense industry, medical industry and nuclear industry. Summarized the status of tools for the software design and requirements management. Analyzed the software design methodology for the safety software of NPP. Development of the design requirements for the requirements tracking and verification system. Development of the background technology to design the prototype tool for the requirements tracking and verification

Survey and assessment of conventional software verification and validation methods

International Nuclear Information System (INIS)

Miller, L.A.; Groundwater, E.; Mirsky, S.M.

1993-04-01

By means of a literature survey, a comprehensive set of methods was identified for the verification and validation of conventional software. The 134 methods so identified were classified according to their appropriateness for various phases of a developmental lifecycle -- requirements, design, and implementation; the last category was subdivided into two, static testing and dynamic testing methods. The methods were then characterized in terms of eight rating factors, four concerning ease-of-use of the methods and four concerning the methods' power to detect defects. Based on these factors, two measurements were developed to permit quantitative comparisons among methods, a Cost-Benefit metric and an Effectiveness Metric. The Effectiveness Metric was further refined to provide three different estimates for each method, depending on three classes of needed stringency of V ampersand V (determined by ratings of a system's complexity and required-integrity). Methods were then rank-ordered for each of the three classes in terms of their overall cost-benefits and effectiveness. The applicability was then assessed of each method for the four identified components of knowledge-based and expert systems, as well as the system as a whole

Concepts of Model Verification and Validation

International Nuclear Information System (INIS)

Thacker, B.H.; Doebling, S.W.; Hemez, F.M.; Anderson, M.C.; Pepin, J.E.; Rodriguez, E.A.

2004-01-01

Model verification and validation (VandV) is an enabling methodology for the development of computational models that can be used to make engineering predictions with quantified confidence. Model VandV procedures are needed by government and industry to reduce the time, cost, and risk associated with full-scale testing of products, materials, and weapon systems. Quantifying the confidence and predictive accuracy of model calculations provides the decision-maker with the information necessary for making high-consequence decisions. The development of guidelines and procedures for conducting a model VandV program are currently being defined by a broad spectrum of researchers. This report reviews the concepts involved in such a program. Model VandV is a current topic of great interest to both government and industry. In response to a ban on the production of new strategic weapons and nuclear testing, the Department of Energy (DOE) initiated the Science-Based Stockpile Stewardship Program (SSP). An objective of the SSP is to maintain a high level of confidence in the safety, reliability, and performance of the existing nuclear weapons stockpile in the absence of nuclear testing. This objective has challenged the national laboratories to develop high-confidence tools and methods that can be used to provide credible models needed for stockpile certification via numerical simulation. There has been a significant increase in activity recently to define VandV methods and procedures. The U.S. Department of Defense (DoD) Modeling and Simulation Office (DMSO) is working to develop fundamental concepts and terminology for VandV applied to high-level systems such as ballistic missile defense and battle management simulations. The American Society of Mechanical Engineers (ASME) has recently formed a Standards Committee for the development of VandV procedures for computational solid mechanics models. The Defense Nuclear Facilities Safety Board (DNFSB) has been a proponent of model

Translating Activity Diagram from Duration Calculus for Modeling of Real-Time Systems and its Formal Verification using UPPAAL and DiVinE

Directory of Open Access Journals (Sweden)

Muhammad Abdul Basit Ur Rehman

2016-01-01

Full Text Available The RTS (Real-Time Systems are widely used in industry, home appliances, life saving systems, aircrafts, and automatic weapons. These systems need more accuracy, safety, and reliability. An accurate graphical modeling and verification of such systems is really challenging. The formal methods made it possible to model such systems with more accuracy. In this paper, we envision a strategy to overcome the inadequacy of SysML (System Modeling Language for modeling and verification of RTS, and illustrate the framework by applying it on a case study of fuel filling machine. We have defined DC (Duration Calculus implementaion based formal semantics to specify the functionality of RTS. The activity diagram in then generated from these semantics. Finally, the graphical model is verified using UPPAAL and DiVinE model checkers for validation of timed and untimed properties with accelerated verification speed. Our results suggest the use of methodology for modeling and verification of large scale real-time systems with reduced verification cost.

Translating activity diagram from duration calculus for modeling of real-time systems and its formal verification using UPPAAL and DiVinE

International Nuclear Information System (INIS)

Rahim, M.A.B.U.; Arif, F.

2016-01-01

The RTS (Real-Time Systems) are widely used in industry, home appliances, life saving systems, aircrafts, and automatic weapons. These systems need more accuracy, safety, and reliability. An accurate graphical modeling and verification of such systems is really challenging. The formal methods made it possible to model such systems with more accuracy. In this paper, we envision a strategy to overcome the inadequacy of SysML (System Modeling Language) for modeling and verification of RTS, and illustrate the framework by applying it on a case study of fuel filling machine. We have defined DC (Duration Calculus) implementation based formal semantics to specify the functionality of RTS. The activity diagram in then generated from these semantics. Finally, the graphical model is verified using UPPAAL and DiVinE model checkers for validation of timed and untimed properties with accelerated verification speed. Our results suggest the use of methodology for modeling and verification of large scale real-time systems with reduced verification cost. (author)

DarcyTools, Version 2.1. Verification and validation

International Nuclear Information System (INIS)

Svensson, Urban

2004-03-01

DarcyTools is a computer code for simulation of flow and transport in porous and/or fractured media. The fractured media in mind is a fractured rock and the porous media the soil cover on the top of the rock; it is hence groundwater flows, which is the class of flows in mind. A number of novel methods and features form the present version of DarcyTools. In the verification studies, these methods are evaluated by comparisons with analytical solutions for idealized situations. The five verification groups, thus reflect the main areas of recent developments. The present report will focus on the Verification and Validation of DarcyTools. Two accompanying reports cover other aspects: - Concepts, Methods, Equations and Demo Simulations. - User's Guide. The objective of this report is to compile all verification and validation studies that have been carried out so far. After some brief introductory sections, all cases will be reported in Appendix A (verification cases) and Appendix B (validation cases)

DarcyTools, Version 2.1. Verification and validation

Energy Technology Data Exchange (ETDEWEB)

Svensson, Urban [Computer-aided Fluid Engineering AB, Norrkoeping (Sweden)

2004-03-01

DarcyTools is a computer code for simulation of flow and transport in porous and/or fractured media. The fractured media in mind is a fractured rock and the porous media the soil cover on the top of the rock; it is hence groundwater flows, which is the class of flows in mind. A number of novel methods and features form the present version of DarcyTools. In the verification studies, these methods are evaluated by comparisons with analytical solutions for idealized situations. The five verification groups, thus reflect the main areas of recent developments. The present report will focus on the Verification and Validation of DarcyTools. Two accompanying reports cover other aspects: - Concepts, Methods, Equations and Demo Simulations. - User's Guide. The objective of this report is to compile all verification and validation studies that have been carried out so far. After some brief introductory sections, all cases will be reported in Appendix A (verification cases) and Appendix B (validation cases)

Present scenery of cuban legislation in the field of legal verification of dosimetric instruments used in radiological protection

International Nuclear Information System (INIS)

Salas G, Walwyn; Morales Monzon, J.A.; Hernandez Blanche, E.

2001-01-01

The main objective of legal metrology is to ensure the public guaranty from the point of view of safety, and the suitable accuracy of the measurements that are made on health, environmental applications, and trade. The International Organization of Legal Metrology included the ionizing radiation field on those for which the use of the verified measuring instruments are suggested. . The paper presents the advances of Cuban legislation in this field, promoted by issue of the Decree-Law 183 of Metrology. As part of such advances, the Cuban standards for verification NC 44:1999 'X and Gamma Radiation Measuring Instruments. Verification methods' is discussed. This standard was elaborated in the Cuban Secondary Standard Dosimetry Laboratory, and it is based on the available relevant international standards. Results from verification service during the year 2000 are also provided.(author)

A Systematic Method for Verification and Validation of Gyrokinetic Microstability Codes

Energy Technology Data Exchange (ETDEWEB)

Bravenec, Ronald [Fourth State Research, Austin, TX (United States)

2017-11-14

My original proposal for the period Feb. 15, 2014 through Feb. 14, 2017 called for an integrated validation and verification effort carried out by myself with collaborators. The validation component would require experimental profile and power-balance analysis. In addition, it would require running the gyrokinetic codes varying the input profiles within experimental uncertainties to seek agreement with experiment before discounting a code as invalidated. Therefore, validation would require a major increase of effort over my previous grant periods which covered only code verification (code benchmarking). Consequently, I had requested full-time funding. Instead, I am being funded at somewhat less than half time (5 calendar months per year). As a consequence, I decided to forego the validation component and to only continue the verification efforts.

An approach to the verification of a fault-tolerant, computer-based reactor safety system: A case study using automated reasoning: Volume 1: Interim report

International Nuclear Information System (INIS)

Chisholm, G.H.; Kljaich, J.; Smith, B.T.; Wojcik, A.S.

1987-01-01

The purpose of this project is to explore the feasibility of automating the verification process for computer systems. The intent is to demonstrate that both the software and hardware that comprise the system meet specified availability and reliability criteria, that is, total design analysis. The approach to automation is based upon the use of Automated Reasoning Software developed at Argonne National Laboratory. This approach is herein referred to as formal analysis and is based on previous work on the formal verification of digital hardware designs. Formal analysis represents a rigorous evaluation which is appropriate for system acceptance in critical applications, such as a Reactor Safety System (RSS). This report describes a formal analysis technique in the context of a case study, that is, demonstrates the feasibility of applying formal analysis via application. The case study described is based on the Reactor Safety System (RSS) for the Experimental Breeder Reactor-II (EBR-II). This is a system where high reliability and availability are tantamount to safety. The conceptual design for this case study incorporates a Fault-Tolerant Processor (FTP) for the computer environment. An FTP is a computer which has the ability to produce correct results even in the presence of any single fault. This technology was selected as it provides a computer-based equivalent to the traditional analog based RSSs. This provides a more conservative design constraint than that imposed by the IEEE Standard, Criteria For Protection Systems For Nuclear Power Generating Stations (ANSI N42.7-1972)

TU-H-CAMPUS-JeP1-02: Fully Automatic Verification of Automatically Contoured Normal Tissues in the Head and Neck

Energy Technology Data Exchange (ETDEWEB)

McCarroll, R [UT MD Anderson Cancer Center, Houston, TX (United States); UT Health Science Center, Graduate School of Biomedical Sciences, Houston, TX (United States); Beadle, B; Yang, J; Zhang, L; Kisling, K; Balter, P; Stingo, F; Nelson, C; Followill, D; Court, L [UT MD Anderson Cancer Center, Houston, TX (United States); Mejia, M [University of Santo Tomas Hospital, Manila, Metro Manila (Philippines)

2016-06-15

Purpose: To investigate and validate the use of an independent deformable-based contouring algorithm for automatic verification of auto-contoured structures in the head and neck towards fully automated treatment planning. Methods: Two independent automatic contouring algorithms [(1) Eclipse’s Smart Segmentation followed by pixel-wise majority voting, (2) an in-house multi-atlas based method] were used to create contours of 6 normal structures of 10 head-and-neck patients. After rating by a radiation oncologist, the higher performing algorithm was selected as the primary contouring method, the other used for automatic verification of the primary. To determine the ability of the verification algorithm to detect incorrect contours, contours from the primary method were shifted from 0.5 to 2cm. Using a logit model the structure-specific minimum detectable shift was identified. The models were then applied to a set of twenty different patients and the sensitivity and specificity of the models verified. Results: Per physician rating, the multi-atlas method (4.8/5 point scale, with 3 rated as generally acceptable for planning purposes) was selected as primary and the Eclipse-based method (3.5/5) for verification. Mean distance to agreement and true positive rate were selected as covariates in an optimized logit model. These models, when applied to a group of twenty different patients, indicated that shifts could be detected at 0.5cm (brain), 0.75cm (mandible, cord), 1cm (brainstem, cochlea), or 1.25cm (parotid), with sensitivity and specificity greater than 0.95. If sensitivity and specificity constraints are reduced to 0.9, detectable shifts of mandible and brainstem were reduced by 0.25cm. These shifts represent additional safety margins which might be considered if auto-contours are used for automatic treatment planning without physician review. Conclusion: Automatically contoured structures can be automatically verified. This fully automated process could be used to

Computational methods assuring nuclear power plant structural integrity and safety: an overview of the recent activities at VTT

International Nuclear Information System (INIS)

Keinaenen, H.; Talja, H.; Rintamaa, R.

1998-01-01

Numerical, simplified engineering and standardised methods are applied in the safety analyses of primary circuit components and reactor pressure vessels. The integrity assessment procedures require input relating both to the steady state and transient loading actual material properties data and precise knowledge of the size and geometry of defects. Current procedures bold extensive information regarding these aspects. It is important to verify the accuracy of the different assessment methods especially in the case of complex structures and loading. The focus of this paper is on the recent results and development of computational fracture assessment methods at VTT Manufacturing Technology. The methods include effective engineering type tools for rapid structural integrity assessments and more sophisticated finite-element based methods. An integrated PC-based program system MASI for engineering fracture analysis is described. A summary of the verification of the methods in computational benchmark analyses and against the results of large scale experiments is presented. (orig.)

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.

Energy Technology Data Exchange (ETDEWEB)

Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

2011-01-01

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

General-Purpose Heat Source Safety Verification Test program: Edge-on flyer plate tests

International Nuclear Information System (INIS)

George, T.G.

1987-03-01

The radioisotope thermoelectric generator (RTG) that will supply power for the Galileo and Ulysses space missions contains 18 General-Purpose Heat Source (GPHS) modules. The GPHS modules provide power by transmitting the heat of 238 Pu α-decay to an array of thermoelectric elements. Each module contains four 238 PuO 2 -fueled clads and generates 250 W(t). Because the possibility of a launch vehicle explosion always exists, and because such an explosion could generate a field of high-energy fragments, the fueled clads within each GPHS module must survive fragment impact. The edge-on flyer plate tests were included in the Safety Verification Test series to provide information on the module/clad response to the impact of high-energy plate fragments. The test results indicate that the edge-on impact of a 3.2-mm-thick, aluminum-alloy (2219-T87) plate traveling at 915 m/s causes the complete release of fuel from capsules contained within a bare GPHS module, and that the threshold velocity sufficient to cause the breach of a bare, simulant-fueled clad impacted by a 3.5-mm-thick, aluminum-alloy (5052-T0) plate is approximately 140 m/s

General-Purpose Heat Source development: Safety Verification Test Program. Bullet/fragment test series

Energy Technology Data Exchange (ETDEWEB)

George, T.G.; Tate, R.E.; Axler, K.M.

1985-05-01

The radioisotope thermoelectric generator (RTG) that will provide power for space missions contains 18 General-Purpose Heat Source (GPHS) modules. Each module contains four /sup 238/PuO/sub 2/-fueled clads and generates 250 W/sub (t)/. Because a launch-pad or post-launch explosion is always possible, we need to determine the ability of GPHS fueled clads within a module to survive fragment impact. The bullet/fragment test series, part of the Safety Verification Test Plan, was designed to provide information on clad response to impact by a compact, high-energy, aluminum-alloy fragment and to establish a threshold value of fragment energy required to breach the iridium cladding. Test results show that a velocity of 555 m/s (1820 ft/s) with an 18-g bullet is at or near the threshold value of fragment velocity that will cause a clad breach. Results also show that an exothermic Ir/Al reaction occurs if aluminum and hot iridium are in contact, a contact that is possible and most damaging to the clad within a narrow velocity range. The observed reactions between the iridium and the aluminum were studied in the laboratory and are reported in the Appendix.

Bibliography for Verification and Validation in Computational Simulation

International Nuclear Information System (INIS)

Oberkampf, W.L.

1998-01-01

A bibliography has been compiled dealing with the verification and validation of computational simulations. The references listed in this bibliography are concentrated in the field of computational fluid dynamics (CFD). However, references from the following fields are also included: operations research, heat transfer, solid dynamics, software quality assurance, software accreditation, military systems, and nuclear reactor safety. This bibliography, containing 221 references, is not meant to be comprehensive. It was compiled during the last ten years in response to the author's interest and research in the methodology for verification and validation. The emphasis in the bibliography is in the following areas: philosophy of science underpinnings, development of terminology and methodology, high accuracy solutions for CFD verification, experimental datasets for CFD validation, and the statistical quantification of model validation. This bibliography should provide a starting point for individual researchers in many fields of computational simulation in science and engineering

Bibliography for Verification and Validation in Computational Simulations

Energy Technology Data Exchange (ETDEWEB)

Oberkampf, W.L.

1998-10-01

A bibliography has been compiled dealing with the verification and validation of computational simulations. The references listed in this bibliography are concentrated in the field of computational fluid dynamics (CFD). However, references from the following fields are also included: operations research, heat transfer, solid dynamics, software quality assurance, software accreditation, military systems, and nuclear reactor safety. This bibliography, containing 221 references, is not meant to be comprehensive. It was compiled during the last ten years in response to the author's interest and research in the methodology for verification and validation. The emphasis in the bibliography is in the following areas: philosophy of science underpinnings, development of terminology and methodology, high accuracy solutions for CFD verification, experimental datasets for CFD validation, and the statistical quantification of model validation. This bibliography should provide a starting point for individual researchers in many fields of computational simulation in science and engineering.

Validation and Verification (V&V) of Safety-Critical Systems Operating Under Off-Nominal Conditions

Science.gov (United States)

Belcastro, Christine M.

2012-01-01

Loss of control (LOC) remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft LOC accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or more often in combination. Hence, there is no single intervention strategy to prevent these accidents. Research is underway at the National Aeronautics and Space Administration (NASA) in the development of advanced onboard system technologies for preventing or recovering from loss of vehicle control and for assuring safe operation under off-nominal conditions associated with aircraft LOC accidents. The transition of these technologies into the commercial fleet will require their extensive validation and verification (V&V) and ultimate certification. The V&V of complex integrated systems poses highly significant technical challenges and is the subject of a parallel research effort at NASA. This chapter summarizes the V&V problem and presents a proposed process that could be applied to complex integrated safety-critical systems developed for preventing aircraft LOC accidents. A summary of recent research accomplishments in this effort is referenced.

Class 1E software verification and validation: Past, present, and future

International Nuclear Information System (INIS)

Persons, W.L.; Lawrence, J.D.

1993-10-01

This paper discusses work in progress that addresses software verification and validation (V ampersand V) as it takes place during the full software life cycle of safety-critical software. The paper begins with a brief overview of the task description and discussion of the historical evolution of software V ampersand V. A new perspective is presented which shows the entire verification and validation process from the viewpoints of a software developer, product assurance engineer, independent V ampersand V auditor, and government regulator. An account of the experience of the field test of the Verification Audit Plan and Report generated from the V ampersand V Guidelines is presented along with sample checklists and lessons learned from the verification audit experience. Then, an approach to automating the V ampersand V Guidelines is introduced. The paper concludes with a glossary and bibliography

Class 1E software verification and validation: Past, present, and future

International Nuclear Information System (INIS)

Persons, W.L.; Lawrence, J.D.

1994-01-01

This paper discusses work in progress that addresses software verification and validation (V ampersand V) as it takes place during the full software life cycle of safety-critical software. The paper begins with a brief overview of the task description and discussion of the historical evolution of software V ampersand V. A new perspective is presented which shows the entire verification and validation process from the viewpoints of a software developer, product assurance engineer, independent V ampersand V auditor, and government regulator. An account of the experience of the field test of the Verification Audit Plan and Report generated from the V ampersand V Guidelines is presented along with sample checklists and lessons learned from the verification audit experience. Then, an approach to automating the V ampersand V Guidelines is introduced. The paper concludes with a glossary and bibliography

Packaged low-level waste verification system

International Nuclear Information System (INIS)

Tuite, K.T.; Winberg, M.; Flores, A.Y.; Killian, E.W.; McIsaac, C.V.

1996-01-01

Currently, states and low-level radioactive waste (LLW) disposal site operators have no method of independently verifying the radionuclide content of packaged LLW that arrive at disposal sites for disposal. At this time, disposal sites rely on LLW generator shipping manifests and accompanying records to insure that LLW received meets the waste acceptance criteria. An independent verification system would provide a method of checking generator LLW characterization methods and help ensure that LLW disposed of at disposal facilities meets requirements. The Mobile Low-Level Waste Verification System (MLLWVS) provides the equipment, software, and methods to enable the independent verification of LLW shipping records to insure that disposal site waste acceptance criteria are being met. The MLLWVS system was developed under a cost share subcontract between WMG, Inc., and Lockheed Martin Idaho Technologies through the Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory (INEL)

Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

International Nuclear Information System (INIS)

Lake, W.H.; Sanders, T.L.; Parks, C.V.

1990-01-01

This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs

Leaf trajectory verification during dynamic intensity modulated radiotherapy using an amorphous silicon flat panel imager

International Nuclear Information System (INIS)

Sonke, Jan-Jakob; Ploeger, Lennert S.; Brand, Bob; Smitsmans, Monique H.P.; Herk, Marcel van

2004-01-01

An independent verification of the leaf trajectories during each treatment fraction improves the safety of IMRT delivery. In order to verify dynamic IMRT with an electronic portal imaging device (EPID), the EPID response should be accurate and fast such that the effect of motion blurring on the detected moving field edge position is limited. In the past, it was shown that the errors in the detected position of a moving field edge determined by a scanning liquid-filled ionization chamber (SLIC) EPID are negligible in clinical practice. Furthermore, a method for leaf trajectory verification during dynamic IMRT was successfully applied using such an EPID. EPIDs based on amorphous silicon (a-Si) arrays are now widely available. Such a-Si flat panel imagers (FPIs) produce portal images with superior image quality compared to other portal imaging systems, but they have not yet been used for leaf trajectory verification during dynamic IMRT. The aim of this study is to quantify the effect of motion distortion and motion blurring on the detection accuracy of a moving field edge for an Elekta iViewGT a-Si FPI and to investigate its applicability for the leaf trajectory verification during dynamic IMRT. We found that the detection error for a moving field edge to be smaller than 0.025 cm at a speed of 0.8 cm/s. Hence, the effect of motion blurring on the detection accuracy of a moving field edge is negligible in clinical practice. Furthermore, the a-Si FPI was successfully applied for the verification of dynamic IMRT. The verification method revealed a delay in the control system of the experimental DMLC that was also found using a SLIC EPID, resulting in leaf positional errors of 0.7 cm at a leaf speed of 0.8 cm/s

Advanced control and instrumentation systems in nuclear power plants. Design, verification and validation

International Nuclear Information System (INIS)

Haapanen, P.

1995-01-01

The Technical Committee Meeting on design, verification and validation of advanced control and instrumentation systems in nuclear power plants was held in Espoo, Finland on 20 - 23 June 1994. The meeting was organized by the International Atomic Energy Agency's (IAEA) International Working Group's (IWG) on Nuclear Power Plant Control and Instrumentation (NPPCI) and on Advanced Technologies for Water Cooled Reactors (ATWR). VTT Automation together with Imatran Voima Oy and Teollisuuden Voima Oy responded about the practical arrangements of the meeting. In total 96 participants from 21 countries and the Agency took part in the meeting and 34 full papers and 8 posters were presented. Following topics were covered in the papers: (1) experience with advanced and digital systems, (2) safety and reliability analysis, (3) advanced digital systems under development and implementation, (4) verification and validation methods and practices, (5) future development trends. (orig.)

FY 1981 report on the results of the verification test on the methanol conversion for oil-fired power plant. Verification test on the environmental safety; 1981 nendo sekiyu karyoku hatsudensho metanoru tenkan tou jissho shiken seika hokokusho. Kankyo anzensei jissho shiken

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-08-01

Assuming the use of methanol which is expected to be promising as petroleum substituting fluid fuel, an investigational study was made on the environmental safety, and the FY 1981 results were summed up. In the study/evaluation of the verification test, conducted were the survey of the results of the studies having been made on toxicity of methanol, working-out of a plan for verification test on the environmental safety of methanol, etc. Moreover, for the purpose of grasping effects of methanol and methanol combustion gas on living organisms, the following were carried out: design and a part of the construction work of facilities in which the test is made for breeding monkey/aquatic animal in the methanol environment, test on its effect on aquatic animal, and purchase of a part of the equipment used for test on its effect on rat/mouse. As to the tests, the following were in the planning stage: toxicity test using macaca on high-concentration (acute)/low-concentration (chronic) inhalation of methanol gas, toxicity test on inhalation of formaldehyde as mock combustion flue gas, test on effects of methanol on fish/shellfish in terms of the fatal concentration/repellent behavior/chronic influence/hindrance of multiplication, etc. (NEDO)

LNG Safety Assessment Evaluation Methods

Energy Technology Data Exchange (ETDEWEB)

Muna, Alice Baca [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-05-01

Sandia National Laboratories evaluated published safety assessment methods across a variety of industries including Liquefied Natural Gas (LNG), hydrogen, land and marine transportation, as well as the US Department of Defense (DOD). All the methods were evaluated for their potential applicability for use in the LNG railroad application. After reviewing the documents included in this report, as well as others not included because of repetition, the Department of Energy (DOE) Hydrogen Safety Plan Checklist is most suitable to be adapted to the LNG railroad application. This report was developed to survey industries related to rail transportation for methodologies and tools that can be used by the FRA to review and evaluate safety assessments submitted by the railroad industry as a part of their implementation plans for liquefied or compressed natural gas storage ( on-board or tender) and engine fueling delivery systems. The main sections of this report provide an overview of various methods found during this survey. In most cases, the reference document is quoted directly. The final section provides discussion and a recommendation for the most appropriate methodology that will allow efficient and consistent evaluations to be made. The DOE Hydrogen Safety Plan Checklist was then revised to adapt it as a methodology for the Federal Railroad Administration’s use in evaluating safety plans submitted by the railroad industry.

Advanced verification methods for OVI security ink

Science.gov (United States)

Coombs, Paul G.; McCaffery, Shaun F.; Markantes, Tom

2006-02-01

OVI security ink +, incorporating OVP security pigment* microflakes, enjoys a history of effective document protection. This security feature provides not only first-line recognition by the person on the street, but also facilitates machine-readability. This paper explores the evolution of OVI reader technology from proof-of-concept to miniaturization. Three different instruments have been built to advance the technology of OVI machine verification. A bench-top unit has been constructed which allows users to automatically verify a multitude of different banknotes and OVI images. In addition, high speed modules were fabricated and tested in a state of the art banknote sorting machine. Both units demonstrate the ability of modern optical components to illuminate and collect light reflected from the interference platelets within OVI ink. Electronic hardware and software convert and process the optical information in milliseconds to accurately determine the authenticity of the security feature. Most recently, OVI ink verification hardware has been miniaturized and simplified providing yet another platform for counterfeit protection. These latest devices provide a tool for store clerks and bank tellers to unambiguously determine the validity of banknotes in the time period it takes the cash drawer to be opened.

NPP Krsko periodic safety review. Safety assessment and analyses

International Nuclear Information System (INIS)

Basic, I.; Spiler, J.; Thaulez, F.

2002-01-01

Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

FY 1983 report on the results of the verification test on the methanol conversion for oil-fired power plant. Part 1. Verification test on the environmental safety; 1983 nendo sekiyu karyoku hatsudensho metanoru tenkan tou jissho shiken seika hokokusho. Kankyo anzensei jissho shiken (Sono 1)

Energy Technology Data Exchange (ETDEWEB)

NONE

1984-03-01

As to the verification test on the environmental safety in the use of methanol as power generation use fuel, the following were summed up: review of the verification test and the interim evaluation, state of implementation of the FY 1983 verification test, study/evaluation of the results of the FY 1983 test, survey of research trends, plan of the FY 1984 verification test, record of the committee, etc. Concerning the interim evaluation, high evaluation was obtained as described below: Testing facilities were constructed as planned at first to make the implementation of various tests possible; Tests were smoothly conducted, and among the acute test using monkey, test on mock flue gas using monkey/rat, test on mutagenicity and test on the effect on aquatic animals, tests using oryzias latipes and abalone on the fatal concentration, avoidance behavior and chronic effect were finished by the end of FY 1983 almost as planned; The long-term inhalation test using monkey and rat/mouse has been smoothly in progress. In the survey of research trends, the paper introduced the outlined literature on the methanol metabolism of monkey, changes in the methanol concentration in blood/urine in the case of drinking methanol by mistake. (NEDO)

Verification of the thermal design of electronic equipment

Energy Technology Data Exchange (ETDEWEB)

Hienonen, R.; Karjalainen, M.; Lankinen, R. [VTT Automation, Espoo (Finland). ProTechno

1997-12-31

The project `Verification of the thermal design of electronic equipment` studied the methodology to be followed in the verification of thermal design of electronic equipment. This project forms part of the `Cool Electronics` research programme funded by TEKES, the Finnish Technology Development Centre. This project was carried out jointly by VTT Automation, Lappeenranta University of Technology, Nokia Research Center and ABB Industry Oy VSD-Technology. The thermal design of electronic equipment has a significant impact on the cost, reliability, tolerance to different environments, selection of components and materials, and ergonomics of the product. This report describes the method for verification of thermal design. It assesses the goals set for thermal design, environmental requirements, technical implementation of the design, thermal simulation and modelling, and design qualification testing and the measurements needed. The verification method covers all packaging levels of electronic equipment from the system level to the electronic component level. The method described in this report can be used as part of the quality system of a corporation. The report includes information about the measurement and test methods needed in the verification process. Some measurement methods for the temperature, flow and pressure of air are described. (orig.) Published in Finnish VTT Julkaisuja 824. 22 refs.

A Verification Method of Inter-Task Cooperation in Embedded Real-time Systems and its Evaluation

Science.gov (United States)

Yoshida, Toshio

In software development process of embedded real-time systems, the design of the task cooperation process is very important. The cooperating process of such tasks is specified by task cooperation patterns. Adoption of unsuitable task cooperation patterns has fatal influence on system performance, quality, and extendibility. In order to prevent repetitive work caused by the shortage of task cooperation performance, it is necessary to verify task cooperation patterns in an early software development stage. However, it is very difficult to verify task cooperation patterns in an early software developing stage where task program codes are not completed yet. Therefore, we propose a verification method using task skeleton program codes and a real-time kernel that has a function of recording all events during software execution such as system calls issued by task program codes, external interrupts, and timer interrupt. In order to evaluate the proposed verification method, we applied it to the software development process of a mechatronics control system.

GRIMHX verification and validation action matrix summary

International Nuclear Information System (INIS)

Trumble, E.F.

1991-12-01

WSRC-RP-90-026, Certification Plan for Reactor Analysis Computer Codes, describes a series of action items to be completed for certification of reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations. Validation and verification of the code is an integral part of this process. This document identifies the work performed and documentation generated to satisfy these action items for the Reactor Physics computer code GRIMHX. Each action item is discussed with the justification for its completion. Specific details of the work performed are not included in this document but are found in the references. The publication of this document signals the validation and verification effort for the GRIMHX code is completed

Verification and Validation of Flight-Critical Systems

Science.gov (United States)

Brat, Guillaume

2010-01-01

For the first time in many years, the NASA budget presented to congress calls for a focused effort on the verification and validation (V&V) of complex systems. This is mostly motivated by the results of the VVFCS (V&V of Flight-Critical Systems) study, which should materialize as a a concrete effort under the Aviation Safety program. This talk will present the results of the study, from requirements coming out of discussions with the FAA and the Joint Planning and Development Office (JPDO) to technical plan addressing the issue, and its proposed current and future V&V research agenda, which will be addressed by NASA Ames, Langley, and Dryden as well as external partners through NASA Research Announcements (NRA) calls. This agenda calls for pushing V&V earlier in the life cycle and take advantage of formal methods to increase safety and reduce cost of V&V. I will present the on-going research work (especially the four main technical areas: Safety Assurance, Distributed Systems, Authority and Autonomy, and Software-Intensive Systems), possible extensions, and how VVFCS plans on grounding the research in realistic examples, including an intended V&V test-bench based on an Integrated Modular Avionics (IMA) architecture and hosted by Dryden.

A simple method for validation and verification of pipettes mounted on automated liquid handlers

DEFF Research Database (Denmark)

Stangegaard, Michael; Hansen, Anders Johannes; Frøslev, Tobias G

2011-01-01

We have implemented a simple, inexpensive, and fast procedure for validation and verification of the performance of pipettes mounted on automated liquid handlers (ALHs) as necessary for laboratories accredited under ISO 17025. A six- or seven-step serial dilution of OrangeG was prepared in quadru......We have implemented a simple, inexpensive, and fast procedure for validation and verification of the performance of pipettes mounted on automated liquid handlers (ALHs) as necessary for laboratories accredited under ISO 17025. A six- or seven-step serial dilution of OrangeG was prepared...... are freely available. In conclusion, we have set up a simple, inexpensive, and fast solution for the continuous validation of ALHs used for accredited work according to the ISO 17025 standard. The method is easy to use for aqueous solutions but requires a spectrophotometer that can read microtiter plates....

Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor

Energy Technology Data Exchange (ETDEWEB)

Setyawan, Daddy, E-mail: d.setyawan@bapeten.go.id [Center for Assessment of Regulatory System and Technology for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia); Rohman, Budi [Licensing Directorate for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia)

2014-09-30

Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

VEG-01: Veggie Hardware Verification Testing

Science.gov (United States)

Massa, Gioia; Newsham, Gary; Hummerick, Mary; Morrow, Robert; Wheeler, Raymond

2013-01-01

The Veggie plant/vegetable production system is scheduled to fly on ISS at the end of2013. Since much of the technology associated with Veggie has not been previously tested in microgravity, a hardware validation flight was initiated. This test will allow data to be collected about Veggie hardware functionality on ISS, allow crew interactions to be vetted for future improvements, validate the ability of the hardware to grow and sustain plants, and collect data that will be helpful to future Veggie investigators as they develop their payloads. Additionally, food safety data on the lettuce plants grown will be collected to help support the development of a pathway for the crew to safely consume produce grown on orbit. Significant background research has been performed on the Veggie plant growth system, with early tests focusing on the development of the rooting pillow concept, and the selection of fertilizer, rooting medium and plant species. More recent testing has been conducted to integrate the pillow concept into the Veggie hardware and to ensure that adequate water is provided throughout the growth cycle. Seed sanitation protocols have been established for flight, and hardware sanitation between experiments has been studied. Methods for shipping and storage of rooting pillows and the development of crew procedures and crew training videos for plant activities on-orbit have been established. Science verification testing was conducted and lettuce plants were successfully grown in prototype Veggie hardware, microbial samples were taken, plant were harvested, frozen, stored and later analyzed for microbial growth, nutrients, and A TP levels. An additional verification test, prior to the final payload verification testing, is desired to demonstrate similar growth in the flight hardware and also to test a second set of pillows containing zinnia seeds. Issues with root mat water supply are being resolved, with final testing and flight scheduled for later in 2013.

Office of River Protection Integrated Safety Management System Phase 1 Verification Corrective Action Plan; FINAL

International Nuclear Information System (INIS)

CLARK, D.L.

1999-01-01

The purpose of this Corrective Action Plan is to demonstrate the OW planned and/or completed actions to implement ISMS as well as prepare for the RPP ISMS Phase II Verification scheduled for August, 1999. This Plan collates implied or explicit ORP actions identified in several key ISMS documents and aligns those actions and responsibilities perceived necessary to appropriately disposition all ISM Phase II preparation activities specific to the ORP. The objective will be to complete or disposition the corrective actions prior to the commencement of the ISMS Phase II Verification. Improvement products/tasks not slated for completion prior to the RPP Phase II verification will be incorporated as corrective actions into the Strategic System Execution Plan (SSEP) Gap Analysis. Many of the business and management systems that were reviewed in the ISMS Phase I verification are being modified to support the ORP transition and are being assessed through the SSEP. The actions and processes identified in the SSEP will support the development of the ORP and continued ISMS implementation as committed to be complete by end of FY-2000

Orion GN&C Fault Management System Verification: Scope And Methodology

Science.gov (United States)

Brown, Denise; Weiler, David; Flanary, Ronald

2016-01-01

In order to ensure long-term ability to meet mission goals and to provide for the safety of the public, ground personnel, and any crew members, nearly all spacecraft include a fault management (FM) system. For a manned vehicle such as Orion, the safety of the crew is of paramount importance. The goal of the Orion Guidance, Navigation and Control (GN&C) fault management system is to detect, isolate, and respond to faults before they can result in harm to the human crew or loss of the spacecraft. Verification of fault management/fault protection capability is challenging due to the large number of possible faults in a complex spacecraft, the inherent unpredictability of faults, the complexity of interactions among the various spacecraft components, and the inability to easily quantify human reactions to failure scenarios. The Orion GN&C Fault Detection, Isolation, and Recovery (FDIR) team has developed a methodology for bounding the scope of FM system verification while ensuring sufficient coverage of the failure space and providing high confidence that the fault management system meets all safety requirements. The methodology utilizes a swarm search algorithm to identify failure cases that can result in catastrophic loss of the crew or the vehicle and rare event sequential Monte Carlo to verify safety and FDIR performance requirements.

Assessment of Automated Measurement and Verification (M&V) Methods

Energy Technology Data Exchange (ETDEWEB)

Granderson, Jessica [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Touzani, Samir [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Custodio, Claudine [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Sohn, Michael [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Fernandes, Samuel [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Jump, David [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2015-07-01

This report documents the application of a general statistical methodology to assess the accuracy of baseline energy models, focusing on its application to Measurement and Verification (M&V) of whole-building energy savings.

Chemical Safety Vulnerability Working Group report. Volume 2

International Nuclear Information System (INIS)

1994-09-01

The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 2 consists of seven appendices containing the following: Tasking memorandums; Project plan for the CSV Review; Field verification guide for the CSV Review; Field verification report, Lawrence Livermore National Lab.; Field verification report, Oak Ridge Reservation; Field verification report, Savannah River Site; and the Field verification report, Hanford Site

Chemical Safety Vulnerability Working Group report. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1994-09-01

The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 2 consists of seven appendices containing the following: Tasking memorandums; Project plan for the CSV Review; Field verification guide for the CSV Review; Field verification report, Lawrence Livermore National Lab.; Field verification report, Oak Ridge Reservation; Field verification report, Savannah River Site; and the Field verification report, Hanford Site.

MESA: Message-Based System Analysis Using Runtime Verification

Science.gov (United States)

Shafiei, Nastaran; Tkachuk, Oksana; Mehlitz, Peter

2017-01-01

In this paper, we present a novel approach and framework for run-time verication of large, safety critical messaging systems. This work was motivated by verifying the System Wide Information Management (SWIM) project of the Federal Aviation Administration (FAA). SWIM provides live air traffic, site and weather data streams for the whole National Airspace System (NAS), which can easily amount to several hundred messages per second. Such safety critical systems cannot be instrumented, therefore, verification and monitoring has to happen using a nonintrusive approach, by connecting to a variety of network interfaces. Due to a large number of potential properties to check, the verification framework needs to support efficient formulation of properties with a suitable Domain Specific Language (DSL). Our approach is to utilize a distributed system that is geared towards connectivity and scalability and interface it at the message queue level to a powerful verification engine. We implemented our approach in the tool called MESA: Message-Based System Analysis, which leverages the open source projects RACE (Runtime for Airspace Concept Evaluation) and TraceContract. RACE is a platform for instantiating and running highly concurrent and distributed systems and enables connectivity to SWIM and scalability. TraceContract is a runtime verication tool that allows for checking traces against properties specified in a powerful DSL. We applied our approach to verify a SWIM service against several requirements.We found errors such as duplicate and out-of-order messages.

Formal Verification -26 ...

Indian Academy of Sciences (India)

by testing of the components and successful testing leads to the software being ... Formal verification is based on formal methods which are mathematically based ..... scenario under which a similar error could occur. There are various other ...

The micro-processor controlled process radiation monitoring system for reactor safety systems

International Nuclear Information System (INIS)

Mizuno, K.; Noguchi, A.; Kumagami, S.; Gotoh, Y.; Kumahara, T.; Arita, S.

1986-01-01

Digital computers are soon expected to be applied to various real-time safety and safety-related systems in nuclear power plants. Hitachi is now engaged in the development of a micro-processor controlled process radiation monitoring system, which operates on digital processing methods employed with a log ratemeter. A newly defined methodology of design and test procedures is being applied as a means of software program verification for these safety systems. Recently implemented micro-processor technology will help to achieve an advanced man-machine interface and highly reliable performance. (author)

Fuel safety criteria technical review - Results of OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria

International Nuclear Information System (INIS)

Hollasky, N.; Valtonen, K.; Hache, G.; Gross, H.; Bakker, K.; Recio, M.; Bart, G.; Zimmermann, M.; Van Doesburg, W.; Killeen, J.; Meyer, R.O.; Speis, T.

2000-01-01

With the advent of advanced fuel and core designs, the adoption of more aggressive operational modes and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a concern if safety margins have remained adequate. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel that was available at that time, mostly with unirradiated specimens. Verification was of course performed as designs progressed in later years, however mostly with the aim to be able to prove that these designs adequately complied with existing criteria, and not to establish new limits. The OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) was therefore given the mandate to technically review the existing fuel safety criteria, focusing on the 'new design' elements (new fuel and core design, cladding materials, manufacturing processes, high burnup, MOX, etc.) introduced by the industry. It should also identify if additional efforts may be required (experimental, analytical) to ensure that the basis for fuel safety criteria is adequate to address the relevant safety issues. In this report, fuel-related criteria are discussed without attempting to categorize them according to event type or risk significance. For each of these 20 criteria, we present a brief description of the criterion as it is used in several applications along with the rationale for having such a criterion. New design elements, such as different cladding materials, higher burnup, and the use of MOX fuels, can affect fuel-related margins and, in some cases, the criteria themselves. Some of the more important effects are mentioned in order to indicate whether the criteria need to be re-evaluated. The discussion may not cover all possible effects, but should be sufficient to identify those criteria that need to be addressed. A summary of these discussions is given in Section 7. As part

Survey and assessment of conventional software verification and validation techniques

International Nuclear Information System (INIS)

Miller, L.A.; Groundwater, E.; Mirsky, S.M.

1993-02-01

Reliable software is required for nuclear power plant applications. Verification and validation (V ampersand V) techniques may be applied during software development to help eliminate errors that can inhibit the proper operation of digital systems and that may cause safety problems. EPRI and the NRC are cosponsoring this investigation to determine the best strategies for V ampersand V of expert system software. The strategy used for a particular system will depend on the complexity of the software and the level of integrity required. This report covers the first task in the investigation of reviewing methods for V ampersand V of conventional software systems and evaluating them for use with expert systems

Enrichment Assay Methods Development for the Integrated Cylinder Verification System

Energy Technology Data Exchange (ETDEWEB)

Smith, Leon E.; Misner, Alex C.; Hatchell, Brian K.; Curtis, Michael M.

2009-10-22

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility's entire product-cylinder inventory. Pacific Northwest National Laboratory (PNNL) is developing a concept to automate the verification of enrichment plant cylinders to enable 100 percent product-cylinder verification and potentially, mass-balance calculations on the facility as a whole (by also measuring feed and tails cylinders). The Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The three main objectives of this FY09 project are summarized here and described in more detail in the report: (1) Develop a preliminary design for a prototype NDA system, (2) Refine PNNL's MCNP models of the NDA system, and (3) Procure and test key pulse-processing components. Progress against these tasks to date, and next steps, are discussed.

Enrichment Assay Methods Development for the Integrated Cylinder Verification System

International Nuclear Information System (INIS)

Smith, Leon E.; Misner, Alex C.; Hatchell, Brian K.; Curtis, Michael M.

2009-01-01

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility's entire product-cylinder inventory. Pacific Northwest National Laboratory (PNNL) is developing a concept to automate the verification of enrichment plant cylinders to enable 100 percent product-cylinder verification and potentially, mass-balance calculations on the facility as a whole (by also measuring feed and tails cylinders). The Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The three main objectives of this FY09 project are summarized here and described in more detail in the report: (1) Develop a preliminary design for a prototype NDA system, (2) Refine PNNL's MCNP models of the NDA system, and (3) Procure and test key pulse-processing components. Progress against these tasks to date, and next steps, are discussed.

International comparison of product certification and verification methods for appliances

Energy Technology Data Exchange (ETDEWEB)

Zhou, Nan [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Romankiewicz, John [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Fridley, David [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Zheng, Nina [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2012-06-01

Enforcement of appliance standards and consumer trust in appliance labeling are important foundations of growing a more energy efficient economy. Product certification and verification increase compliance rates which in turn increase both energy savings and consumer trust. This paper will serve two purposes: 1) to review international practices for product certification and verification as they relate to the enforcement of standards and labeling programs in the U.S., E.U., Australia, Japan, Canada, and China; and 2) to make recommendations for China to implement improved certification processes related to their mandatory standards and labeling program such as to increase compliance rates and energy savings potential.

International Comparison of Product Certification and Verification Methods for Appliances

Energy Technology Data Exchange (ETDEWEB)

Zhou, Nan [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Romankiewicz, John [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Fridley, David [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Zheng, Nina [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2012-06-01

Enforcement of appliance standards and consumer trust in appliance labeling are important foundations of growing a more energy efficient economy. Product certification and verification increase compliance rates which in turn increase both energy savings and consumer trust. This paper will serve two purposes: 1) to review international practices for product certification and verification as they relate to the enforcement of standards and labeling programs in the U.S., E.U., Australia, Japan, Canada, and China; and 2) to make recommendations for China to implement improved certification processes related to their mandatory standards and labeling program such as to increase compliance rates and energy savings potential.

AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

Energy Technology Data Exchange (ETDEWEB)

Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

2016-05-15

Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

Verification and Validation of a Coordinate Transformation Method in Axisymmetric Transient Magnetics.

Energy Technology Data Exchange (ETDEWEB)

Ashcraft, C. Chace [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Niederhaus, John Henry [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Robinson, Allen C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-01-29

We present a verification and validation analysis of a coordinate-transformation-based numerical solution method for the two-dimensional axisymmetric magnetic diffusion equation, implemented in the finite-element simulation code ALEGRA. The transformation, suggested by Melissen and Simkin, yields an equation set perfectly suited for linear finite elements and for problems with large jumps in material conductivity near the axis. The verification analysis examines transient magnetic diffusion in a rod or wire in a very low conductivity background by first deriving an approximate analytic solution using perturbation theory. This approach for generating a reference solution is shown to be not fully satisfactory. A specialized approach for manufacturing an exact solution is then used to demonstrate second-order convergence under spatial refinement and tem- poral refinement. For this new implementation, a significant improvement relative to previously available formulations is observed. Benefits in accuracy for computed current density and Joule heating are also demonstrated. The validation analysis examines the circuit-driven explosion of a copper wire using resistive magnetohydrodynamics modeling, in comparison to experimental tests. The new implementation matches the accuracy of the existing formulation, with both formulations capturing the experimental burst time and action to within approximately 2%.

Fast film dosimetry calibration method for IMRT treatment plan verification

International Nuclear Information System (INIS)

Schwob, N.; Wygoda, A.

2004-01-01

Intensity-Modulated Radiation Therapy (IMRT) treatments are delivered dynamically and as so, require routinely performed verification measurements [1]. Radiographic film dosimetry is a well-adapted method for integral measurements of dynamic treatments fields, with some drawbacks related to the known problems of dose calibration of films. Classically, several films are exposed to increasing doses, and a Net Optical Density (N.O.D) vs. dose sensitometric curve (S.C.) is generated. In order to speed up the process, some authors have developed a method based on the irradiation of a single film with a non-uniform pattern of O.D., delivered with a dynamic MLC. However, this curve still needs to be calibrated to dose by the means of measurements in a water phantom. It is recommended to make a new calibration for every series of measurements, in order to avoid the processing quality dependence of the film response. These frequent measurements are very time consuming. We developed a simple method for quick dose calibration of films, including a check of the accuracy of the calibration curve obtained

Formal verification of industrial control systems

CERN Multimedia

CERN. Geneva

2015-01-01

Verification of critical software is a high priority but a challenging task for industrial control systems. For many kinds of problems, testing is not an efficient method. Formal methods, such as model checking appears to be an appropriate complementary method. However, it is not common to use model checking in industry yet, as this method needs typically formal methods expertise and huge computing power. In the EN-ICE-PLC section, we are working on a [methodology][1] and a tool ([PLCverif][2]) to overcome these challenges and to integrate formal verification in the development process of our PLC-based control systems. [1]: http://cern.ch/project-plc-formalmethods [2]: http://cern.ch/plcverif

Software for computers in the safety systems of nuclear power stations

International Nuclear Information System (INIS)

1987-08-01

This standard includes the safety actuation systems, the safety system support features and the protection systems. The standard provides requirements for each stage of software generation, including design, development, qualification and operation as well as the documentation for each stage of the software generation for the purpose of achieving highly reliable software. The principles applied in developing these requirements include: Best available practice; top-down design methods; modularity; verification of each phase; clear documentation; auditable documents and validation testing. (orig./HP)

Quantifying system safety: A comparison of the SBOAT & Safety Barrier Manager tools

OpenAIRE

Hansen, Zaza Nadja Lee; Duijm, Nijs Jan; Markert, Frank; Herbert, Luke Thomas

2015-01-01

This paper presents two software tools for analyzing safety risks, SBOAT (Stochastic BPMN Optimisation and Analysis Tool) and SBM (SafetyBarrierManagerr). SBOAT employs principles from stochastic model checking to allow for the quantitative verification of workflows. SBM supports the creation of valid safety-barrier diagrams and allows the quantitative analysis of the probability of all possible end states of the barrier diagram, i.e. the outcomes if one or several of the barriers fail to per...

Survey of Verification and Validation Techniques for Small Satellite Software Development

Science.gov (United States)

Jacklin, Stephen A.

2015-01-01

The purpose of this paper is to provide an overview of the current trends and practices in small-satellite software verification and validation. This document is not intended to promote a specific software assurance method. Rather, it seeks to present an unbiased survey of software assurance methods used to verify and validate small satellite software and to make mention of the benefits and value of each approach. These methods include simulation and testing, verification and validation with model-based design, formal methods, and fault-tolerant software design with run-time monitoring. Although the literature reveals that simulation and testing has by far the longest legacy, model-based design methods are proving to be useful for software verification and validation. Some work in formal methods, though not widely used for any satellites, may offer new ways to improve small satellite software verification and validation. These methods need to be further advanced to deal with the state explosion problem and to make them more usable by small-satellite software engineers to be regularly applied to software verification. Last, it is explained how run-time monitoring, combined with fault-tolerant software design methods, provides an important means to detect and correct software errors that escape the verification process or those errors that are produced after launch through the effects of ionizing radiation.

Formal verification of Simulink/Stateflow diagrams a deductive approach

CERN Document Server

Zhan, Naijun; Zhao, Hengjun

2017-01-01

This book presents a state-of-the-art technique for formal verification of continuous-time Simulink/Stateflow diagrams, featuring an expressive hybrid system modelling language, a powerful specification logic and deduction-based verification approach, and some impressive, realistic case studies. Readers will learn the HCSP/HHL-based deductive method and the use of corresponding tools for formal verification of Simulink/Stateflow diagrams. They will also gain some basic ideas about fundamental elements of formal methods such as formal syntax and semantics, and especially the common techniques applied in formal modelling and verification of hybrid systems. By investigating the successful case studies, readers will realize how to apply the pure theory and techniques to real applications, and hopefully will be inspired to start to use the proposed approach, or even develop their own formal methods in their future work.

Evaluating software for safety systems in nuclear power plants

International Nuclear Information System (INIS)

Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.

1994-01-01

In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper

Neighbors Based Discriminative Feature Difference Learning for Kinship Verification

DEFF Research Database (Denmark)

Duan, Xiaodong; Tan, Zheng-Hua

2015-01-01

In this paper, we present a discriminative feature difference learning method for facial image based kinship verification. To transform feature difference of an image pair to be discriminative for kinship verification, a linear transformation matrix for feature difference between an image pair...... than the commonly used feature concatenation, leading to a low complexity. Furthermore, there is no positive semi-definitive constrain on the transformation matrix while there is in metric learning methods, leading to an easy solution for the transformation matrix. Experimental results on two public...... databases show that the proposed method combined with a SVM classification method outperforms or is comparable to state-of-the-art kinship verification methods. © Springer International Publishing AG, Part of Springer Science+Business Media...

Verification station for Sandia/Rockwell Plutonium Protection system

International Nuclear Information System (INIS)

Nicholson, N.; Hastings, R.D.; Henry, C.N.; Millegan, D.R.

1979-04-01

A verification station has been designed to confirm the presence of plutonium within a container module. These container modules [about 13 cm (5 in.) in diameter and 23 cm (9 in.) high] hold sealed food-pack cans containing either plutonium oxide or metal and were designed by Sandia Laboratories to provide security and continuous surveillance and safety. After the plutonium is placed in the container module, it is closed with a solder seal. The verification station discussed here is used to confirm the presence of plutonium in the container module before it is placed in a carousel-type storage array inside the plutonium storage vault. This measurement represents the only technique that uses nuclear detectors in the plutonium protection system

Safety assessment of geologic repositories for nuclear waste

International Nuclear Information System (INIS)

Bartlett, J.W.; Burkholder, H.C.; Winegardner, W.K.

1977-01-01

Consideration of geologic isolation for final disposition of radioactive wastes has led to the need for evaluation of the safety of the concept. Such evaluations require consideration of factors not encountered in conventional risk analysis: consequences at times and places far removed from the repository site; indirect, complex, and alternative pathways between the waste and the point of potential consequences; a highly limited data base; and limited opportunity for experimental verification of results. R and D programs to provide technical safety evaluations are under way. Three methods are being considered for the probabilistic aspects of the evaluations: fault tree analysis, repository simulation analysis, and system stability analysis. Nuclide transport models, currently in a relatively advanced state of development, are used to evaluate consequences of postulated loss of geologic isolation. This paper outlines the safety assessment methods, unique features of the assessment problem that affect selection of methods and reliability of results, and available results. It also discusses potential directions for future work

Status of safety issues at licensed power plants

International Nuclear Information System (INIS)

1991-06-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. This report, the third volume of a three-volume series, addresses the status of generic safety issues (GSIs) at licensed plants. Volume 1 addressed the status of Three Mile Island Action Plan requirements and was published in March 1991. Volume 2 addressed the status of implementation and verification of unresolved safety issues and was published in May 1991. The annual NUREG report will combine these three areas in a single volume to be published in late 1991. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The purpose of this report is to provide a comprehensive description of the status of implementation and verification of the 34 GSIs and sub-issues that have been resolved by the NRC and involve implementation of an action or actions by licensees. This NUREG report also serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until a request for action by licensees is issued by NRC. 3 figs., 6 tabs

Verification of the coupled space-angle adaptivity algorithm for the finite element-spherical harmonics method via the method of manufactured solutions

International Nuclear Information System (INIS)

Park, H.; De Oliveira, C. R. E.

2007-01-01

This paper describes the verification of the recently developed space-angle self-adaptive algorithm for the finite element-spherical harmonics method via the Method of Manufactured Solutions. This method provides a simple, yet robust way for verifying the theoretical properties of the adaptive algorithm and interfaces very well with the underlying second-order, even-parity transport formulation. Simple analytic solutions in both spatial and angular variables are manufactured to assess the theoretical performance of the a posteriori error estimates. The numerical results confirm reliability of the developed space-angle error indicators. (authors)

Verification of fluid-structure-interaction algorithms through the method of manufactured solutions for actuator-line applications

Science.gov (United States)

Vijayakumar, Ganesh; Sprague, Michael

2017-11-01

Demonstrating expected convergence rates with spatial- and temporal-grid refinement is the ``gold standard'' of code and algorithm verification. However, the lack of analytical solutions and generating manufactured solutions presents challenges for verifying codes for complex systems. The application of the method of manufactured solutions (MMS) for verification for coupled multi-physics phenomena like fluid-structure interaction (FSI) has only seen recent investigation. While many FSI algorithms for aeroelastic phenomena have focused on boundary-resolved CFD simulations, the actuator-line representation of the structure is widely used for FSI simulations in wind-energy research. In this work, we demonstrate the verification of an FSI algorithm using MMS for actuator-line CFD simulations with a simplified structural model. We use a manufactured solution for the fluid velocity field and the displacement of the SMD system. We demonstrate the convergence of both the fluid and structural solver to second-order accuracy with grid and time-step refinement. This work was funded by the U.S. Department of Energy, Office of Energy Efficiency and Renewable Energy, Wind Energy Technologies Office, under Contract No. DE-AC36-08-GO28308 with the National Renewable Energy Laboratory.

Comparison of the methods for determination of calibration and verification intervals of measuring devices

Directory of Open Access Journals (Sweden)

Toteva Pavlina

2017-01-01

Full Text Available The paper presents different determination and optimisation methods for verification intervals of technical devices for monitoring and measurement based on the requirements of some widely used international standards, e.g. ISO 9001, ISO/IEC 17020, ISO/IEC 17025 etc., maintained by various organizations implementing measuring devices in practice. Comparative analysis of the reviewed methods is conducted in terms of opportunities for assessing the adequacy of interval(s for calibration of measuring devices and their optimisation accepted by an organization – an extension or reduction depending on the obtained results. The advantages and disadvantages of the reviewed methods are discussed, and recommendations for their applicability are provided.

Integrated knowledge base tool for acquisition and verification of NPP alarm systems

International Nuclear Information System (INIS)

Park, Joo Hyun; Seong, Poong Hyun

1998-01-01

Knowledge acquisition and knowledge base verification are important activities in developing knowledge-based systems such as alarm processing systems. In this work, we developed the integrated tool, for knowledge acquisition and verification of NPP alarm processing systems, by using G2 tool. The tool integrates document analysis method and ECPN matrix analysis method, for knowledge acquisition and knowledge verification, respectively. This tool enables knowledge engineers to perform their tasks from knowledge acquisition to knowledge verification consistently

Content analysis of age verification, purchase and delivery methods of internet e-cigarette vendors, 2013 and 2014.

Science.gov (United States)

Williams, Rebecca S; Derrick, Jason; Liebman, Aliza Kate; LaFleur, Kevin; Ribisl, Kurt M

2018-05-01

Identify the population of internet e-cigarette vendors (IEVs) and conduct content analyses of their age verification, purchase and delivery methods in 2013 and 2014. We used multiple sources to identify IEV websites, primarily complex search algorithms scanning more than 180 million websites. In 2013, we manually screened 32 446 websites, identifying 980 IEVs, selecting the 281 most popular for content analysis. This methodology yielded 31 239 websites for screening in 2014, identifying 3096 IEVs, with 283 selected for content analysis. The proportion of vendors that sold online-only, with no retail store, dropped significantly from 2013 (74.7%) to 2014 (64.3%) (ponline age verification services (7.1% in 2013 and 8.5% in 2014), driving licences (1.8% in 2013 and 7.4% in 2014, ponline e-cigarette sales are needed, including strict age and identity verification requirements. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

Property-based Code Slicing for Efficient Verification of OSEK/VDX Operating Systems

Directory of Open Access Journals (Sweden)

Mingyu Park

2012-12-01

Full Text Available Testing is a de-facto verification technique in industry, but insufficient for identifying subtle issues due to its optimistic incompleteness. On the other hand, model checking is a powerful technique that supports comprehensiveness, and is thus suitable for the verification of safety-critical systems. However, it generally requires more knowledge and cost more than testing. This work attempts to take advantage of both techniques to achieve integrated and efficient verification of OSEK/VDX-based automotive operating systems. We propose property-based environment generation and model extraction techniques using static code analysis, which can be applied to both model checking and testing. The technique is automated and applied to an OSEK/VDX-based automotive operating system, Trampoline. Comparative experiments using random testing and model checking for the verification of assertions in the Trampoline kernel code show how our environment generation and abstraction approach can be utilized for efficient fault-detection.

FPGA Design and Verification Procedure for Nuclear Power Plant MMIS

Energy Technology Data Exchange (ETDEWEB)

Lee, Dongil; Yoo, Kawnwoo; Ryoo, Kwangki [Hanbat National Univ., Daejeon (Korea, Republic of)

2013-05-15

In this paper, it is shown that it is possible to ensure reliability by performing the steps of the verification based on the FPGA development methodology, to ensure the safety of application to the NPP MMIS of the FPGA run along the step. Currently, the PLC (Programmable Logic Controller) which is being developed is composed of the FPGA (Field Programmable Gate Array) and CPU (Central Processing Unit). As the importance of the FPGA in the NPP (Nuclear Power Plant) MMIS (Man-Machine Interface System) has been increasing than before, the research on the verification of the FPGA has being more and more concentrated recently.

Preparation of Input Deck to analyze the Nuclear Power Plant for the Use of Regulatory Verification

International Nuclear Information System (INIS)

Kang, Doo Hyuk; Kim, Hyung Seok; Suh, Jae Seung; Ahn, Seung Hoon

2009-01-01

The objectives of this paper are to make out the input deck that analyzes a nuclear power plant for the use of regulatory verification and to produce its calculation note. We have been maintained the input deck of T/H safety codes used in existing domestic reactors to ensure independent and accurate regulatory verification for the thermal-hydraulic safety analysis in domestic NPPs. This paper is mainly divided into two steps: first step is to compare existing input deck to the calculation note in order to verify the consistency. Next step is to model 3-dimensional reactor pressure vessel using MULTID component instead of the 1D existing input deck

Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Hollnagel, E. [Linkoeping Univ. (Sweden). Dept. of Computer and Information Science; Gauthereau, V. [Linkoeping Univ. (Sweden). Dept. of Industrial Engineering

2001-06-01

This report contains the findings from the first phase of a study on safety during outage and restart of nuclear power plants. Operational Readiness Verification (ORV) - in Swedish called Driftklarhetsverifiering (DKV) - refers to the test and verification activities that are necessary to ensure that plant systems are able to provide their required functions when needed - more concretely that all plant systems are in their correct functional state when the plant is restarted after an outage period. The concrete background for this work is that nine ORV related incidents were reported in Sweden between July 1995 and October 1998. The work reported here comprised a literature survey of research relevant for ORV issues, and an assessment of the present situation at Swedish NPPs with respect to ORV. The literature survey was primarily aimed at research related to NPPs, but also looked at domains where similar problems have occurred, such as maintenance in commercial aviation. The survey looked specifically for organisational and MTO aspects relevant to the present situation in Swedish NPPs. One finding was that ORV should be seen as an integral part of maintenance, rather than as a separate activity. Another, that there is a characteristic distribution of error modes for maintenance and ORV, with many sequence errors and omissions, rather than a set of unique error modes. An international study further showed that there are important differences in how procedures are used, and in the balance between decentralisation and centralisation. Several studies also suggested that ORV could usefully be described as a barrier system in relation to the flow of work, for instance using the following five stages: (1) preventive actions during maintenance/outage, (2) post-test after completion of work, (3) pre-test before start-up, (4) the start-up sequence itself, and (5) preventive actions during power operation - possibly including automatic safety systems. In the field survey

Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

International Nuclear Information System (INIS)

Hollnagel, E.; Gauthereau, V.

2001-06-01

This report contains the findings from the first phase of a study on safety during outage and restart of nuclear power plants. Operational Readiness Verification (ORV) - in Swedish called Driftklarhetsverifiering (DKV) - refers to the test and verification activities that are necessary to ensure that plant systems are able to provide their required functions when needed - more concretely that all plant systems are in their correct functional state when the plant is restarted after an outage period. The concrete background for this work is that nine ORV related incidents were reported in Sweden between July 1995 and October 1998. The work reported here comprised a literature survey of research relevant for ORV issues, and an assessment of the present situation at Swedish NPPs with respect to ORV. The literature survey was primarily aimed at research related to NPPs, but also looked at domains where similar problems have occurred, such as maintenance in commercial aviation. The survey looked specifically for organisational and MTO aspects relevant to the present situation in Swedish NPPs. One finding was that ORV should be seen as an integral part of maintenance, rather than as a separate activity. Another, that there is a characteristic distribution of error modes for maintenance and ORV, with many sequence errors and omissions, rather than a set of unique error modes. An international study further showed that there are important differences in how procedures are used, and in the balance between decentralisation and centralisation. Several studies also suggested that ORV could usefully be described as a barrier system in relation to the flow of work, for instance using the following five stages: (1) preventive actions during maintenance/outage, (2) post-test after completion of work, (3) pre-test before start-up, (4) the start-up sequence itself, and (5) preventive actions during power operation - possibly including automatic safety systems. In the field survey

Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Hollnagel, E [Linkoeping Univ. (Sweden). Dept. of Computer and Information Science; Gauthereau, V [Linkoeping Univ. (Sweden). Dept. of Industrial Engineering

2001-06-01

This report contains the findings from the first phase of a study on safety during outage and restart of nuclear power plants. Operational Readiness Verification (ORV) - in Swedish called Driftklarhetsverifiering (DKV) - refers to the test and verification activities that are necessary to ensure that plant systems are able to provide their required functions when needed - more concretely that all plant systems are in their correct functional state when the plant is restarted after an outage period. The concrete background for this work is that nine ORV related incidents were reported in Sweden between July 1995 and October 1998. The work reported here comprised a literature survey of research relevant for ORV issues, and an assessment of the present situation at Swedish NPPs with respect to ORV. The literature survey was primarily aimed at research related to NPPs, but also looked at domains where similar problems have occurred, such as maintenance in commercial aviation. The survey looked specifically for organisational and MTO aspects relevant to the present situation in Swedish NPPs. One finding was that ORV should be seen as an integral part of maintenance, rather than as a separate activity. Another, that there is a characteristic distribution of error modes for maintenance and ORV, with many sequence errors and omissions, rather than a set of unique error modes. An international study further showed that there are important differences in how procedures are used, and in the balance between decentralisation and centralisation. Several studies also suggested that ORV could usefully be described as a barrier system in relation to the flow of work, for instance using the following five stages: (1) preventive actions during maintenance/outage, (2) post-test after completion of work, (3) pre-test before start-up, (4) the start-up sequence itself, and (5) preventive actions during power operation - possibly including automatic safety systems. In the field survey

Advanced verification topics

CERN Document Server

Bhattacharya, Bishnupriya; Hall, Gary; Heaton, Nick; Kashai, Yaron; Khan Neyaz; Kirshenbaum, Zeev; Shneydor, Efrat

2011-01-01

The Accellera Universal Verification Methodology (UVM) standard is architected to scale, but verification is growing and in more than just the digital design dimension. It is growing in the SoC dimension to include low-power and mixed-signal and the system integration dimension to include multi-language support and acceleration. These items and others all contribute to the quality of the SOC so the Metric-Driven Verification (MDV) methodology is needed to unify it all into a coherent verification plan. This book is for verification engineers and managers familiar with the UVM and the benefits it brings to digital verification but who also need to tackle specialized tasks. It is also written for the SoC project manager that is tasked with building an efficient worldwide team. While the task continues to become more complex, Advanced Verification Topics describes methodologies outside of the Accellera UVM standard, but that build on it, to provide a way for SoC teams to stay productive and profitable.

Methods and Effects of Safety Enhancement in Korean PSR

International Nuclear Information System (INIS)

Kim, Young Gab; Park, Jong Woon

2009-01-01

Periodic Safety Review (PSR) is a comprehensive study on a nuclear power plant safety, taking into account aspects such as operational history, ageing, safety analyses and advances in code and standards since the time of construction. In Korea, PSRs have been performed for 20 units and have been effectively used to obtain an overall view of actual plant safety to determine reasonable and practical modifications that should be made in order to obtain a higher level of safety approaching that of modern plants. Among many safety enhancements achieved from Korean PSRs, new safety analyses are the important methods to confirm plant safety by increasing safety margin for specific safety issues. Methods and effects of safety enhancements applied in Korean PSRs are reviewed in this paper in light of new safety analyses to obtain additional safety margins

Verification tests for remote controlled inspection system in nuclear power plants

International Nuclear Information System (INIS)

Kohno, Tadaaki

1986-01-01

Following the increase of nuclear power plants, the total radiation exposure dose accompanying inspection and maintenance works tended to increase. Japan Power Engineering and Inspection Corp. carried out the verification test of a practical power reactor automatic inspection system from November, 1981, to March, 1986, and in this report, the state of having carried out this verification test is described. The objects of the verification test were the equipment which is urgently required for reducing radiation exposure dose, the possibility of realization of which is high, and which is important for ensuring the safety and reliability of plants, that is, an automatic ultrasonic flaw detector for the welded parts of bend pipes, an automatic disassembling and inspection system for control rod driving mechanism, a fuel automatic inspection system, and automatic decontaminating equipments for steam generator water chambers, primary system crud and radioactive gas in coolant. The results of the verification test of these equipments were judged as satisfactory, therefore, the application to actual plants is possible. (Kako, I.)

Safety of Research Reactors. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2017-01-01

This Safety Requirements publication establishes requirements for all main areas of safety for research reactors, with particular emphasis on requirements for design and operation. It explains the safety objectives and concepts that form the basis for safety and safety assessment for all stages in the lifetime of a research reactor. Technical and administrative requirements for the safety of new research reactors are established in accordance with these objectives and concepts, and they are to be applied to the extent practicable for existing research reactors. The safety requirements established in this publication for the management of safety and regulatory supervision apply to site evaluation, design, manufacturing, construction, commissioning, operation (including utilization and modification), and planning for decommissioning of research reactors (including critical assemblies and subcritical assemblies). The publication is intended for use by regulatory bodies and other organizations with responsibilities in these areas and in safety analysis, verification and review, and the provision of technical support.

37 CFR 262.7 - Verification of royalty payments.

Science.gov (United States)

2010-07-01

... Designated Agent have agreed as to proper verification methods. (b) Frequency of verification. A Copyright Owner or a Performer may conduct a single audit of the Designated Agent upon reasonable notice and... COPYRIGHT ARBITRATION ROYALTY PANEL RULES AND PROCEDURES RATES AND TERMS FOR CERTAIN ELIGIBLE...

Multi-resistance strategy for viral diseases and short hairpin RNA verification method in pigs

Directory of Open Access Journals (Sweden)

Jong-nam Oh

2018-04-01

Full Text Available Objective Foot and mouth disease (FMD and porcine reproductive and respiratory syndrome (PRRS are major diseases that interrupt porcine production. Because they are viral diseases, vaccinations are of only limited effectiveness in preventing outbreaks. To establish an alternative multi-resistant strategy against FMD virus (FMDV and PRRS virus (PRRSV, the present study introduced two genetic modification techniques to porcine cells. Methods First, cluster of differentiation 163 (CD163, the PRRSV viral receptor, was edited with the clustered regularly interspaced short palindromic repeats-CRISPR-associated protein 9 technique. The CD163 gene sequences of edited cells and control cells differed. Second, short hairpin RNA (shRNAs were integrated into the cells. The shRNAs, targeting the 3D gene of FMDV and the open reading frame 7 (ORF7 gene of PRRSV, were transferred into fibroblasts. We also developed an in vitro shRNA verification method with a target gene expression vector. Results shRNA activity was confirmed in vitro with vectors that expressed the 3D and ORF7 genes in the cells. Cells containing shRNAs showed lower transcript levels than cells with only the expression vectors. The shRNAs were integrated into CD163-edited cells to combine the two techniques, and the viral genes were suppressed in these cells. Conclusion We established a multi-resistant strategy against viral diseases and an in vitro shRNA verification method.

Design and Verification of Critical Pressurised Windows for Manned Spaceflight

Science.gov (United States)

Lamoure, Richard; Busto, Lara; Novo, Francisco; Sinnema, Gerben; Leal, Mendes M.

2014-06-01

The Window Design for Manned Spaceflight (WDMS) project was tasked with establishing the state-of-art and explore possible improvements to the current structural integrity verification and fracture control methodologies for manned spacecraft windows.A critical review of the state-of-art in spacecraft window design, materials and verification practice was conducted. Shortcomings of the methodology in terms of analysis, inspection and testing were identified. Schemes for improving verification practices and reducing conservatism whilst maintaining the required safety levels were then proposed.An experimental materials characterisation programme was defined and carried out with the support of the 'Glass and Façade Technology Research Group', at the University of Cambridge. Results of the sample testing campaign were analysed, post-processed and subsequently applied to the design of a breadboard window demonstrator.Two Fused Silica glass window panes were procured and subjected to dedicated analyses, inspection and testing comprising both qualification and acceptance programmes specifically tailored to the objectives of the activity.Finally, main outcomes have been compiled into a Structural Verification Guide for Pressurised Windows in manned spacecraft, incorporating best practices and lessons learned throughout this project.

The safety implications of emerging software paradigms

International Nuclear Information System (INIS)

Suski, G.J.; Persons, W.L.; Johnson, G.L.

1994-10-01

This paper addresses some of the emerging software paradigms that may be used in developing safety-critical software applications. Paradigms considered in this paper include knowledge-based systems, neural networks, genetic algorithms, and fuzzy systems. It presents one view of the software verification and validation activities that should be associated with each paradigm. The paper begins with a discussion of the historical evolution of software verification and validation. Next, a comparison is made between the verification and validation processes used for conventional and emerging software systems. Several verification and validation issues for the emerging paradigms are discussed and some specific research topics are identified. This work is relevant for monitoring and control at nuclear power plants

Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

International Nuclear Information System (INIS)

2014-01-01

The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

Test-driven verification/validation of model transformations

Institute of Scientific and Technical Information of China (English)

László LENGYEL; Hassan CHARAF

2015-01-01

Why is it important to verify/validate model transformations? The motivation is to improve the quality of the trans-formations, and therefore the quality of the generated software artifacts. Verified/validated model transformations make it possible to ensure certain properties of the generated software artifacts. In this way, verification/validation methods can guarantee different requirements stated by the actual domain against the generated/modified/optimized software products. For example, a verified/ validated model transformation can ensure the preservation of certain properties during the model-to-model transformation. This paper emphasizes the necessity of methods that make model transformation verified/validated, discusses the different scenarios of model transformation verification and validation, and introduces the principles of a novel test-driven method for verifying/ validating model transformations. We provide a solution that makes it possible to automatically generate test input models for model transformations. Furthermore, we collect and discuss the actual open issues in the field of verification/validation of model transformations.

Development of Advanced Verification and Validation Procedures and Tools for the Certification of Learning Systems in Aerospace Applications

Science.gov (United States)

Jacklin, Stephen; Schumann, Johann; Gupta, Pramod; Richard, Michael; Guenther, Kurt; Soares, Fola

2005-01-01

Adaptive control technologies that incorporate learning algorithms have been proposed to enable automatic flight control and vehicle recovery, autonomous flight, and to maintain vehicle performance in the face of unknown, changing, or poorly defined operating environments. In order for adaptive control systems to be used in safety-critical aerospace applications, they must be proven to be highly safe and reliable. Rigorous methods for adaptive software verification and validation must be developed to ensure that control system software failures will not occur. Of central importance in this regard is the need to establish reliable methods that guarantee convergent learning, rapid convergence (learning) rate, and algorithm stability. This paper presents the major problems of adaptive control systems that use learning to improve performance. The paper then presents the major procedures and tools presently developed or currently being developed to enable the verification, validation, and ultimate certification of these adaptive control systems. These technologies include the application of automated program analysis methods, techniques to improve the learning process, analytical methods to verify stability, methods to automatically synthesize code, simulation and test methods, and tools to provide on-line software assurance.

Quantitative analysis of patient-specific dosimetric IMRT verification

International Nuclear Information System (INIS)

Budgell, G J; Perrin, B A; Mott, J H L; Fairfoul, J; Mackay, R I

2005-01-01

Patient-specific dosimetric verification methods for IMRT treatments are variable, time-consuming and frequently qualitative, preventing evidence-based reduction in the amount of verification performed. This paper addresses some of these issues by applying a quantitative analysis parameter to the dosimetric verification procedure. Film measurements in different planes were acquired for a series of ten IMRT prostate patients, analysed using the quantitative parameter, and compared to determine the most suitable verification plane. Film and ion chamber verification results for 61 patients were analysed to determine long-term accuracy, reproducibility and stability of the planning and delivery system. The reproducibility of the measurement and analysis system was also studied. The results show that verification results are strongly dependent on the plane chosen, with the coronal plane particularly insensitive to delivery error. Unexpectedly, no correlation could be found between the levels of error in different verification planes. Longer term verification results showed consistent patterns which suggest that the amount of patient-specific verification can be safely reduced, provided proper caution is exercised: an evidence-based model for such reduction is proposed. It is concluded that dose/distance to agreement (e.g., 3%/3 mm) should be used as a criterion of acceptability. Quantitative parameters calculated for a given criterion of acceptability should be adopted in conjunction with displays that show where discrepancies occur. Planning and delivery systems which cannot meet the required standards of accuracy, reproducibility and stability to reduce verification will not be accepted by the radiotherapy community

The safety issues of medical robotics

Energy Technology Data Exchange (ETDEWEB)

Fei Baowei; Ng, W.S.; Chauhan, Sunita; Kwoh, Chee Keong

2001-08-01

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 modules adaptable for 3D ultrasound image-guided interstitial laser coagulation, radiation seed implantation, laser resection, and electrical resection of the prostate. Safety was always the key issue in the building of the robot. The HISIC strategies were adopted for safety enhancement in mechanical, electrical and software design. The initial test on URObot showed that HISIC had the potential ability to improve the safety of the system. Further safety experiments are being conducted in our laboratory.

The safety issues of medical robotics

International Nuclear Information System (INIS)

Fei Baowei; Ng, W.S.; Chauhan, Sunita; Kwoh, Chee Keong

2001-01-01

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 modules adaptable for 3D ultrasound image-guided interstitial laser coagulation, radiation seed implantation, laser resection, and electrical resection of the prostate. Safety was always the key issue in the building of the robot. The HISIC strategies were adopted for safety enhancement in mechanical, electrical and software design. The initial test on URObot showed that HISIC had the potential ability to improve the safety of the system. Further safety experiments are being conducted in our laboratory

Application of FMEA method in railway signalling projects

Directory of Open Access Journals (Sweden)

Szmel Dariusz

2017-06-01

Full Text Available The article presents the FMEA method application, which is relevant in verification of design of two separated railway signalling systems. The efficiency of the method at the stage of the design was discussed. The method was identified as an important element of safety management process and as safety analysis method, which is included in the Safety Case and is applied for the sake of safety arguments and its assessment. Safety process management comprises several phases and appropriate actions, linked with each other in the way to create safety life cycle consistent with system life cycle. The safety case is a set of documents demonstrating that the product is compliant with defined safety requirements including analysis that indicates the correctness of the design and the correct reaction of the system to the failures, with appropriate and requested fail-safe reaction. It is necessary that railway signalling system should fulfil SIL4 requirement and remain safe in case of occurrence any kind of single failure of the equipment considered as possible.

Watershed safety and quality control by safety threshold method

Science.gov (United States)

Da-Wei Tsai, David; Mengjung Chou, Caroline; Ramaraj, Rameshprabu; Liu, Wen-Cheng; Honglay Chen, Paris

2014-05-01

Taiwan was warned as one of the most dangerous countries by IPCC and the World Bank. In such an exceptional and perilous island, we would like to launch the strategic research of land-use management on the catastrophe prevention and environmental protection. This study used the watershed management by "Safety Threshold Method" to restore and to prevent the disasters and pollution on island. For the deluge prevention, this study applied the restoration strategy to reduce total runoff which was equilibrium to 59.4% of the infiltration each year. For the sediment management, safety threshold management could reduce the sediment below the equilibrium of the natural sediment cycle. In the water quality issues, the best strategies exhibited the significant total load reductions of 10% in carbon (BOD5), 15% in nitrogen (nitrate) and 9% in phosphorus (TP). We found out the water quality could meet the BOD target by the 50% peak reduction with management. All the simulations demonstrated the safety threshold method was helpful to control the loadings within the safe range of disasters and environmental quality. Moreover, from the historical data of whole island, the past deforestation policy and the mistake economic projects were the prime culprits. Consequently, this study showed a practical method to manage both the disasters and pollution in a watershed scale by the land-use management.

46 CFR 61.40-6 - Periodic safety tests.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Periodic safety tests. 61.40-6 Section 61.40-6 Shipping... INSPECTIONS Design Verification and Periodic Testing of Vital System Automation § 61.40-6 Periodic safety tests. (a) Periodic Safety tests must demonstrate the proper operation of the primary and alternate...

The awareness of employees in safety culture through the improved nuclear safety culture evaluation method

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Ga; Sung, Chan Ho; Jung, Yeon Sub [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2012-10-15

After the Chernobyl nuclear accident in 1986, nuclear safety culture terminology was at first introduced emphasizing the importance of employees' attitude and organizational safety. The concept of safety culture was spread by INSAG 4 published in 1991. From that time, IAEA had provided the service of ASCOT for the safety culture assessment. However, many people still are thinking that safety culture is abstract and is not clear. It is why the systematic and reliable assessment methodology was not developed. Assessing safety culture is to identify what is the basic assumption for any organization to accept unconsciously. Therefore, it is very difficult to reach a meaningful conclusion by a superficial investigation alone. KHNP had been doing the safety culture assessment which was based on ASCOT methodology every 2 years. And this result had contributed to improving safety culture. But this result could not represent the level of organization's safety culture due to the limitation of method. So, KHNP has improved the safety culture method by benchmarking the over sea assessment techniques in 2011. The effectiveness of this improved methodology was validated through a pilot assessment. In this paper, the level of employees' safety culture awareness was analyzed by the improved method and reviewed what is necessary for the completeness and objectivity of the nuclear safety culture assessment methodology.

The awareness of employees in safety culture through the improved nuclear safety culture evaluation method

International Nuclear Information System (INIS)

Kim, Young Ga; Sung, Chan Ho; Jung, Yeon Sub

2012-01-01

After the Chernobyl nuclear accident in 1986, nuclear safety culture terminology was at first introduced emphasizing the importance of employees' attitude and organizational safety. The concept of safety culture was spread by INSAG 4 published in 1991. From that time, IAEA had provided the service of ASCOT for the safety culture assessment. However, many people still are thinking that safety culture is abstract and is not clear. It is why the systematic and reliable assessment methodology was not developed. Assessing safety culture is to identify what is the basic assumption for any organization to accept unconsciously. Therefore, it is very difficult to reach a meaningful conclusion by a superficial investigation alone. KHNP had been doing the safety culture assessment which was based on ASCOT methodology every 2 years. And this result had contributed to improving safety culture. But this result could not represent the level of organization's safety culture due to the limitation of method. So, KHNP has improved the safety culture method by benchmarking the over sea assessment techniques in 2011. The effectiveness of this improved methodology was validated through a pilot assessment. In this paper, the level of employees' safety culture awareness was analyzed by the improved method and reviewed what is necessary for the completeness and objectivity of the nuclear safety culture assessment methodology

Shield verification and validation action matrix summary

International Nuclear Information System (INIS)

Boman, C.

1992-02-01

WSRC-RP-90-26, Certification Plan for Reactor Analysis Computer Codes, describes a series of action items to be completed for certification of reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations. Validation and verification are integral part of the certification process. This document identifies the work performed and documentation generated to satisfy these action items for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system, it is not certification of the complete SHIELD system. Complete certification will follow at a later date. Each action item is discussed with the justification for its completion. Specific details of the work performed are not included in this document but can be found in the references. The validation and verification effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system computer code is completed

Automated Formal Verification for PLC Control Systems

CERN Multimedia

Fernández Adiego, Borja

2014-01-01

Programmable Logic Controllers (PLCs) are widely used devices used in industrial control systems. Ensuring that the PLC software is compliant with its specification is a challenging task. Formal verification has become a recommended practice to ensure the correctness of the safety-critical software. However, these techniques are still not widely applied in industry due to the complexity of building formal models, which represent the system and the formalization of requirement specifications. We propose a general methodology to perform automated model checking of complex properties expressed in temporal logics (e.g. CTL, LTL) on PLC programs. This methodology is based on an Intermediate Model (IM), meant to transform PLC programs written in any of the languages described in the IEC 61131-3 standard (ST, IL, etc.) to different modeling languages of verification tools. This approach has been applied to CERN PLC programs validating the methodology.

Applying Formal Verification Techniques to Ambient Assisted Living Systems

Science.gov (United States)

Benghazi, Kawtar; Visitación Hurtado, María; Rodríguez, María Luisa; Noguera, Manuel

This paper presents a verification approach based on timed traces semantics and MEDISTAM-RT [1] to check the fulfillment of non-functional requirements, such as timeliness and safety, and assure the correct functioning of the Ambient Assisted Living (AAL) systems. We validate this approach by its application to an Emergency Assistance System for monitoring people suffering from cardiac alteration with syncope.

Secure optical verification using dual phase-only correlation

International Nuclear Information System (INIS)

Liu, Wei; Liu, Shutian; Zhang, Yan; Xie, Zhenwei; Liu, Zhengjun

2015-01-01

We introduce a security-enhanced optical verification system using dual phase-only correlation based on a novel correlation algorithm. By employing a nonlinear encoding, the inherent locks of the verification system are obtained in real-valued random distributions, and the identity keys assigned to authorized users are designed as pure phases. The verification process is implemented in two-step correlation, so only authorized identity keys can output the discriminate auto-correlation and cross-correlation signals that satisfy the reset threshold values. Compared with the traditional phase-only-correlation-based verification systems, a higher security level against counterfeiting and collisions are obtained, which is demonstrated by cryptanalysis using known attacks, such as the known-plaintext attack and the chosen-plaintext attack. Optical experiments as well as necessary numerical simulations are carried out to support the proposed verification method. (paper)

Verification test report on a solar heating and hot water system

Science.gov (United States)

1978-01-01

Information is provided on the development, qualification and acceptance verification of commercial solar heating and hot water systems and components. The verification includes the performances, the efficiences and the various methods used, such as similarity, analysis, inspection, test, etc., that are applicable to satisfying the verification requirements.

Cleanup Verification Package for the 118-F-3, Minor Construction Burial Ground

International Nuclear Information System (INIS)

Appel, M.J.

2007-01-01

This cleanup verification package documents completion of remedial action for the 118-F-3, Minor Construction Burial Ground waste site. This site was an open field covered with cobbles, with no vegetation growing on the surface. The site received irradiated reactor parts that were removed during conversion of the 105-F Reactor from the Liquid 3X to the Ball 3X Project safety systems and received mostly vertical safety rod thimbles and step plugs

Safety of nuclear fuel cycle facilities. Safety requirements

International Nuclear Information System (INIS)

2008-01-01

This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

Heavy water physical verification in power plants

International Nuclear Information System (INIS)

Morsy, S.; Schuricht, V.; Beetle, T.; Szabo, E.

1986-01-01

This paper is a report on the Agency experience in verifying heavy water inventories in power plants. The safeguards objectives and goals for such activities are defined in the paper. The heavy water is stratified according to the flow within the power plant, including upgraders. A safeguards scheme based on a combination of records auditing, comparing records and reports, and physical verification has been developed. This scheme has elevated the status of heavy water safeguards to a level comparable to nuclear material safeguards in bulk facilities. It leads to attribute and variable verification of the heavy water inventory in the different system components and in the store. The verification methods include volume and weight determination, sampling and analysis, non-destructive assay (NDA), and criticality check. The analysis of the different measurement methods and their limits of accuracy are discussed in the paper

Technical challenges for dismantlement verification

International Nuclear Information System (INIS)

Olinger, C.T.; Stanbro, W.D.; Johnston, R.G.; Nakhleh, C.W.; Dreicer, J.S.

1997-01-01

In preparation for future nuclear arms reduction treaties, including any potential successor treaties to START I and II, the authors have been examining possible methods for bilateral warhead dismantlement verification. Warhead dismantlement verification raises significant challenges in the political, legal, and technical arenas. This discussion will focus on the technical issues raised by warhead arms controls. Technical complications arise from several sources. These will be discussed under the headings of warhead authentication, chain-of-custody, dismantlement verification, non-nuclear component tracking, component monitoring, and irreversibility. The authors will discuss possible technical options to address these challenges as applied to a generic dismantlement and disposition process, in the process identifying limitations and vulnerabilities. They expect that these considerations will play a large role in any future arms reduction effort and, therefore, should be addressed in a timely fashion

Biometric Technologies and Verification Systems

CERN Document Server

Vacca, John R

2007-01-01

Biometric Technologies and Verification Systems is organized into nine parts composed of 30 chapters, including an extensive glossary of biometric terms and acronyms. It discusses the current state-of-the-art in biometric verification/authentication, identification and system design principles. It also provides a step-by-step discussion of how biometrics works; how biometric data in human beings can be collected and analyzed in a number of ways; how biometrics are currently being used as a method of personal identification in which people are recognized by their own unique corporal or behavior

Temporal Specification and Verification of Real-Time Systems.

Science.gov (United States)

1991-08-30

of concrete real - time systems can be modeled adequately. Specification: We present two conservative extensions of temporal logic that allow for the...logic. We present both model-checking algorithms for the automatic verification of finite-state real - time systems and proof methods for the deductive verification of real - time systems .

Wavelet-based verification of the quantitative precipitation forecast

Science.gov (United States)

Yano, Jun-Ichi; Jakubiak, Bogumil

2016-06-01

This paper explores the use of wavelets for spatial verification of quantitative precipitation forecasts (QPF), and especially the capacity of wavelets to provide both localization and scale information. Two 24-h forecast experiments using the two versions of the Coupled Ocean/Atmosphere Mesoscale Prediction System (COAMPS) on 22 August 2010 over Poland are used to illustrate the method. Strong spatial localizations and associated intermittency of the precipitation field make verification of QPF difficult using standard statistical methods. The wavelet becomes an attractive alternative, because it is specifically designed to extract spatially localized features. The wavelet modes are characterized by the two indices for the scale and the localization. Thus, these indices can simply be employed for characterizing the performance of QPF in scale and localization without any further elaboration or tunable parameters. Furthermore, spatially-localized features can be extracted in wavelet space in a relatively straightforward manner with only a weak dependence on a threshold. Such a feature may be considered an advantage of the wavelet-based method over more conventional "object" oriented verification methods, as the latter tend to represent strong threshold sensitivities. The present paper also points out limits of the so-called "scale separation" methods based on wavelets. Our study demonstrates how these wavelet-based QPF verifications can be performed straightforwardly. Possibilities for further developments of the wavelet-based methods, especially towards a goal of identifying a weak physical process contributing to forecast error, are also pointed out.

A method of knowledge base verification and validation for nuclear power plants expert systems

International Nuclear Information System (INIS)

Kwon, Il Won

1996-02-01

The adoption of expert systems mainly as operator supporting systems is becoming increasingly popular as the control algorithms of system become more and more sophisticated and complicated. As a result of this popularity, a large number of expert systems are developed. The nature of expert systems, however, requires that they be verified and validated carefully and that detailed methodologies for their development be devised. Therefore, it is widely noted that assuring the reliability of expert systems is very important, especially in nuclear industry, and it is also recognized that the process of verification and validation is an essential part of reliability assurance for these systems. Research and practices have produced numerous methods for expert system verification and validation (V and V) that suggest traditional software and system approaches to V and V. However, many approaches and methods for expert system V and V are partial, unreliable, and not uniform. The purpose of this paper is to present a new approach to expert system V and V, based on Petri nets, providing a uniform model. We devise and suggest an automated tool, called COKEP (Checker Of Knowledge base using Extended Petri net), for checking incorrectness, inconsistency, and incompleteness in a knowledge base. We also suggest heuristic analysis for validation process to show that the reasoning path is correct

Safety review on unit testing of safety system software of nuclear power plant

International Nuclear Information System (INIS)

Liu Le; Zhang Qi

2013-01-01

Software unit testing has an important place in the testing of safety system software of nuclear power plants, and in the wider scope of the verification and validation. It is a comprehensive, systematic process, and its documentation shall meet the related requirements. When reviewing software unit testing, attention should be paid to the coverage of software safety requirements, the coverage of software internal structure, and the independence of the work. (authors)

SIMMER as a safety analysis tool

International Nuclear Information System (INIS)

Smith, L.L.; Bell, C.R.; Bohl, W.R.; Bott, T.F.; Dearing, J.F.; Luck, L.B.

1982-01-01

SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Energy Technology Data Exchange (ETDEWEB)

Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

2017-02-15

Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

Role of FFTF in assessing structural feedbacks and inherent safety of LMR's

International Nuclear Information System (INIS)

Padilla, A.; Omberg, R.P.; O'Dell, L.D.; Harris, R.A.; Nguyen, D.H.; Waltar, A.E.

1985-03-01

The possibility of developing reactor designs with inherent safety characteristics sufficient to provide ''walk away'' safety is receiving additional emphasis in the LMR program. A key element in this effort is the recognition that LMR's possess safety characteristics above and beyond those employed in past safety review processes. Some of these additional safety characteristics are due to reactivity feedback effects caused by small structural movements during hypothetical severe design transients. The effect of these characteristics upon the behavior of the FFTF under such transients has been assessed and is discussed in this paper. The paper also presents a preliminary test matrix which might allow experimental verification of the structural reactivity feedback effects. Such experimental verification should be very useful to innovative designers seeking to optimize inherent safety. 8 refs., 1 fig., 2 tabs

Solution verification, goal-oriented adaptive methods for stochastic advection–diffusion problems

KAUST Repository

Almeida, Regina C.

2010-08-01

A goal-oriented analysis of linear, stochastic advection-diffusion models is presented which provides both a method for solution verification as well as a basis for improving results through adaptation of both the mesh and the way random variables are approximated. A class of model problems with random coefficients and source terms is cast in a variational setting. Specific quantities of interest are specified which are also random variables. A stochastic adjoint problem associated with the quantities of interest is formulated and a posteriori error estimates are derived. These are used to guide an adaptive algorithm which adjusts the sparse probabilistic grid so as to control the approximation error. Numerical examples are given to demonstrate the methodology for a specific model problem. © 2010 Elsevier B.V.

Solution verification, goal-oriented adaptive methods for stochastic advection–diffusion problems

KAUST Repository

Almeida, Regina C.; Oden, J. Tinsley

2010-01-01

A goal-oriented analysis of linear, stochastic advection-diffusion models is presented which provides both a method for solution verification as well as a basis for improving results through adaptation of both the mesh and the way random variables are approximated. A class of model problems with random coefficients and source terms is cast in a variational setting. Specific quantities of interest are specified which are also random variables. A stochastic adjoint problem associated with the quantities of interest is formulated and a posteriori error estimates are derived. These are used to guide an adaptive algorithm which adjusts the sparse probabilistic grid so as to control the approximation error. Numerical examples are given to demonstrate the methodology for a specific model problem. © 2010 Elsevier B.V.

An Introduction of Behavior-Based Safety Program in Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Yong Hee; Lim, Hyeon Kyo

2011-01-01

There are many methods and approaches for a human error assessment that is valuable for investigating the causes of undesirable events and counter-plans to prevent their recurrence in the nuclear power plants (NPPs). There is behavior-based safety refers to the process of using a proactive approach to safety and health management. It either focuses on risk of behaviors that can lead to an injury, or on safe behaviors that can contribute to injury prevention. Early applications of behavior based safety included the construction and manufacturing industries, but today behavior based safety is applied to a wide variety of industries and service lines. This behavior based safety program can offer a set of significant human error countermeasures to be considered for human error in NPPs as well as other fields of industry. The current methods for the human error prevention in NPPs are several techniques such as Self-Check, Peer Check, Concurrent Verification, 3-way Communication, etc. However, it is not enough to grasp the whole human error problems in operations because the things are needed in fields are a behavior technique not a simple knowledge. Therefore, we applied a behavior based safety program on the current methods

12 CFR 715.8 - Requirements for verification of accounts and passbooks.

Science.gov (United States)

2010-01-01

...' share and loan accounts; (2) Statistical method. A sampling method which provides for: (i) Random... CREDIT UNIONS SUPERVISORY COMMITTEE AUDITS AND VERIFICATIONS § 715.8 Requirements for verification of... jurisdiction in which the credit union is principally located, the auditor may choose among the sampling...

Verification and validation for waste disposal models

International Nuclear Information System (INIS)

1987-07-01

A set of evaluation criteria has been developed to assess the suitability of current verification and validation techniques for waste disposal methods. A survey of current practices and techniques was undertaken and evaluated using these criteria with the items most relevant to waste disposal models being identified. Recommendations regarding the most suitable verification and validation practices for nuclear waste disposal modelling software have been made

Research on key technology of the verification system of steel rule based on vision measurement

Science.gov (United States)

Jia, Siyuan; Wang, Zhong; Liu, Changjie; Fu, Luhua; Li, Yiming; Lu, Ruijun

2018-01-01

The steel rule plays an important role in quantity transmission. However, the traditional verification method of steel rule based on manual operation and reading brings about low precision and low efficiency. A machine vison based verification system of steel rule is designed referring to JJG1-1999-Verificaiton Regulation of Steel Rule [1]. What differentiates this system is that it uses a new calibration method of pixel equivalent and decontaminates the surface of steel rule. Experiments show that these two methods fully meet the requirements of the verification system. Measuring results strongly prove that these methods not only meet the precision of verification regulation, but also improve the reliability and efficiency of the verification system.

Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

International Nuclear Information System (INIS)

Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

1994-07-01

The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

A methodology for the rigorous verification of plasma simulation codes

Science.gov (United States)

Riva, Fabio

2016-10-01

The methodology used to assess the reliability of numerical simulation codes constitutes the Verification and Validation (V&V) procedure. V&V is composed by two separate tasks: the verification, which is a mathematical issue targeted to assess that the physical model is correctly solved, and the validation, which determines the consistency of the code results, and therefore of the physical model, with experimental data. In the present talk we focus our attention on the verification, which in turn is composed by the code verification, targeted to assess that a physical model is correctly implemented in a simulation code, and the solution verification, that quantifies the numerical error affecting a simulation. Bridging the gap between plasma physics and other scientific domains, we introduced for the first time in our domain a rigorous methodology for the code verification, based on the method of manufactured solutions, as well as a solution verification based on the Richardson extrapolation. This methodology was applied to GBS, a three-dimensional fluid code based on a finite difference scheme, used to investigate the plasma turbulence in basic plasma physics experiments and in the tokamak scrape-off layer. Overcoming the difficulty of dealing with a numerical method intrinsically affected by statistical noise, we have now generalized the rigorous verification methodology to simulation codes based on the particle-in-cell algorithm, which are employed to solve Vlasov equation in the investigation of a number of plasma physics phenomena.

Models and methods for hot spot safety work

DEFF Research Database (Denmark)

Vistisen, Dorte

2002-01-01

Despite the fact that millions DKK each year are spent on improving roadsafety in Denmark, funds for traffic safety are limited. It is therefore vital to spend the resources as effectively as possible. This thesis is concerned with the area of traffic safety denoted "hot spot safety work", which...... is the task of improving road safety through alterations of the geometrical and environmental characteristics of the existing road network. The presently applied models and methods in hot spot safety work on the Danish road network were developed about two decades ago, when data was more limited and software...... and statistical methods less developed. The purpose of this thesis is to contribute to improving "State of the art" in Denmark. Basis for the systematic hot spot safety work are the models describing the variation in accident counts on the road network. In the thesis hierarchical models disaggregated on time...

Independent assembly technology for DCS safety equipment in nuclear power plant

International Nuclear Information System (INIS)

Hao Aixia

2014-01-01

A independent assembly technology of identification and verification was proposed, which included special process, wiring process and verification process. The safety reliability and practicability of the proposed technology were verified according to the application in FirmSys assembly implemented by China Techenergy Co., Ltd. (author)

Operation safety of control systems. Principles and methods

International Nuclear Information System (INIS)

Aubry, J.F.; Chatelet, E.

2008-01-01

This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)

Procedural method for the development of scenarios in the post-closure phase. Report on the working package 1. Development of the international status of science and technology concerning methods and tools for operational and long-term safety cases; Vorgehensweise bei der Entwicklung von Szenarien fuer die Nachverschlussphase. Bericht zum Arbeitspaket 1. Weiterentwicklung des internationalen Stands von Wissenschaft und Technik zu Methoden und Werkzeugen fuer Betriebs- und Langzeitsicherheitsnachweise

Energy Technology Data Exchange (ETDEWEB)

Beuth, Thomas; Mayer, Kim-Marisa

2016-09-15

The report on the procedural method for the development of scenarios in the post-closure phase covers the following topics: development of scenarios and derivation of calculation cases, approaches for verification of derived scenarios, human penetration in a final repository (including national and international regulations and guidelines and safety standards).

Status of safety issues at licensed power plants

International Nuclear Information System (INIS)

1991-05-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. This report, the second volume of a three-volume series, addresses the status of unresolved safety issues (USIs) at licensed plants. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The purpose of this report is to provide a comprehensive description of the status of implementation and verification of the 27 safety issues designated as USIs and to make this information available to other interested parties, including the public. A corollary purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants. 3 figs., 4 tabs

Software safety analysis on the model specified by NuSCR and SMV input language at requirements phase of software development life cycle using SMV

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun

2005-01-01

Safety-critical software process is composed of development process, verification and validation (V and V) process and safety analysis process. Safety analysis process has been often treated as an additional process and not found in a conventional software process. But software safety analysis (SSA) is required if software is applied to a safety system, and the SSA shall be performed independently for the safety software through software development life cycle (SDLC). Of all the phases in software development, requirements engineering is generally considered to play the most critical role in determining the overall software quality. NASA data demonstrate that nearly 75% of failures found in operational software were caused by errors in the requirements. The verification process in requirements phase checks the correctness of software requirements specification, and the safety analysis process analyzes the safety-related properties in detail. In this paper, the method for safety analysis at requirements phase of software development life cycle using symbolic model verifier (SMV) is proposed. Hazard is discovered by hazard analysis and in other to use SMV for the safety analysis, the safety-related properties are expressed by computation tree logic (CTL)

RELAP-7 Software Verification and Validation Plan: Requirements Traceability Matrix (RTM) Part 1 – Physics and numerical methods

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Joon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoo, Jun Soo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

This INL plan comprehensively describes the Requirements Traceability Matrix (RTM) on main physics and numerical method of the RELAP-7. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7.

A Synthesized Framework for Formal Verification of Computing Systems

Directory of Open Access Journals (Sweden)

Nikola Bogunovic

2003-12-01

Full Text Available Design process of computing systems gradually evolved to a level that encompasses formal verification techniques. However, the integration of formal verification techniques into a methodical design procedure has many inherent miscomprehensions and problems. The paper explicates the discrepancy between the real system implementation and the abstracted model that is actually used in the formal verification procedure. Particular attention is paid to the seamless integration of all phases of the verification procedure that encompasses definition of the specification language and denotation and execution of conformance relation between the abstracted model and its intended behavior. The concealed obstacles are exposed, computationally expensive steps identified and possible improvements proposed.

Verification of statistical method CORN for modeling of microfuel in the case of high grain concentration

Energy Technology Data Exchange (ETDEWEB)

Chukbar, B. K., E-mail: bchukbar@mail.ru [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

Two methods of modeling a double-heterogeneity fuel are studied: the deterministic positioning and the statistical method CORN of the MCU software package. The effect of distribution of microfuel in a pebble bed on the calculation results is studied. The results of verification of the statistical method CORN for the cases of the microfuel concentration up to 170 cm{sup –3} in a pebble bed are presented. The admissibility of homogenization of the microfuel coating with the graphite matrix is studied. The dependence of the reactivity on the relative location of fuel and graphite spheres in a pebble bed is found.

Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues

International Nuclear Information System (INIS)

1992-12-01

This report is to provide a comprehensive description of the implementation and verification status of Three Mile Island (TMI) Action Plan requirements, safety issues designated as Unresolved Safety Issues (USIs), Generic Safety Issues(GSIs), and other Multiplant Actions (MPAs) that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

Verification and quality control of routine hematology analyzers.

Science.gov (United States)

Vis, J Y; Huisman, A

2016-05-01

Verification of hematology analyzers (automated blood cell counters) is mandatory before new hematology analyzers may be used in routine clinical care. The verification process consists of several items which comprise among others: precision, accuracy, comparability, carryover, background and linearity throughout the expected range of results. Yet, which standard should be met or which verification limit be used is at the discretion of the laboratory specialist. This paper offers practical guidance on verification and quality control of automated hematology analyzers and provides an expert opinion on the performance standard that should be met by the contemporary generation of hematology analyzers. Therefore (i) the state-of-the-art performance of hematology analyzers for complete blood count parameters is summarized, (ii) considerations, challenges, and pitfalls concerning the development of a verification plan are discussed, (iii) guidance is given regarding the establishment of reference intervals, and (iv) different methods on quality control of hematology analyzers are reviewed. © 2016 John Wiley & Sons Ltd.

A usability review of a model checker VIS for the verification of NPP I and C system safety software

International Nuclear Information System (INIS)

Son, H. S.; Kwon, K. C.

2002-01-01

This paper discusses the usability of a model checker VIS in the verification of safety software of NPP I and C systems. The software development environment exemplified in this paper is for PLC and ESF-CCS which are being developed in KNICS project. In this environment, STATEMATE is used in requirement analysis and design phases. PLC is expected to be implemented using C language and an assembly language because it has many interfaces with hardware like CPU, I/O devices, communication devices. ESF-CCS is supposed to be developed in terms of PLC programming languages which are defined in IEC 61131-3 standard. In this case, VIS proved to be very useful through the review. We are also able to expect greater usability of VIS if we further develop the techniques for code abstraction and automatic translation from code to verilog, which is the input of VIS

Conducting organizational safety reviews - requirements, methods and experience

International Nuclear Information System (INIS)

Reiman, T.; Oedewald, P.; Wahlstroem, B.; Rollenhagen, C.; Kahlbom, U.

2008-03-01

Organizational safety reviews are part of the safety management process of power plants. They are typically performed after major reorganizations, significant incidents or according to specified review programs. Organizational reviews can also be a part of a benchmarking between organizations that aims to improve work practices. Thus, they are important instruments in proactive safety management and safety culture. Most methods that have been used for organizational reviews are based more on practical considerations than a sound scientific theory of how various organizational or technical issues influence safety. Review practices and methods also vary considerably. The objective of this research is to promote understanding on approaches used in organizational safety reviews as well as to initiate discussion on criteria and methods of organizational assessment. The research identified a set of issues that need to be taken into account when planning and conducting organizational safety reviews. Examples of the issues are definition of appropriate criteria for evaluation, the expertise needed in the assessment and the organizational motivation for conducting the assessment. The study indicates that organizational safety assessments involve plenty of issues and situations where choices have to be made regarding what is considered valid information and a balance has to be struck between focus on various organizational phenomena. It is very important that these choices are based on a sound theoretical framework and that these choices can later be evaluated together with the assessment findings. The research concludes that at its best, the organizational safety reviews can be utilised as a source of information concerning the changing vulnerabilities and the actual safety performance of the organization. In order to do this, certain basic organizational phenomena and assessment issues have to be acknowledged and considered. The research concludes with recommendations on

Conducting organizational safety reviews - requirements, methods and experience

Energy Technology Data Exchange (ETDEWEB)

Reiman, T.; Oedewald, P.; Wahlstroem, B. [Technical Research Centre of Finland, VTT (Finland); Rollenhagen, C. [Royal Institute of Technology, KTH, (Sweden); Kahlbom, U. [RiskPilot (Sweden)

2008-03-15

Organizational safety reviews are part of the safety management process of power plants. They are typically performed after major reorganizations, significant incidents or according to specified review programs. Organizational reviews can also be a part of a benchmarking between organizations that aims to improve work practices. Thus, they are important instruments in proactive safety management and safety culture. Most methods that have been used for organizational reviews are based more on practical considerations than a sound scientific theory of how various organizational or technical issues influence safety. Review practices and methods also vary considerably. The objective of this research is to promote understanding on approaches used in organizational safety reviews as well as to initiate discussion on criteria and methods of organizational assessment. The research identified a set of issues that need to be taken into account when planning and conducting organizational safety reviews. Examples of the issues are definition of appropriate criteria for evaluation, the expertise needed in the assessment and the organizational motivation for conducting the assessment. The study indicates that organizational safety assessments involve plenty of issues and situations where choices have to be made regarding what is considered valid information and a balance has to be struck between focus on various organizational phenomena. It is very important that these choices are based on a sound theoretical framework and that these choices can later be evaluated together with the assessment findings. The research concludes that at its best, the organizational safety reviews can be utilised as a source of information concerning the changing vulnerabilities and the actual safety performance of the organization. In order to do this, certain basic organizational phenomena and assessment issues have to be acknowledged and considered. The research concludes with recommendations on

Verification of analysis methods for predicting the behaviour of seismically isolated nuclear structures. Final report of a co-ordinated research project 1996-1999

International Nuclear Information System (INIS)

2002-06-01

This report is a summary of the work performed under a co-ordinated research project (CRP) entitled Verification of Analysis Methods for Predicting the Behaviour of Seismically isolated Nuclear Structures. The project was organized by the IAEA on the recommendation of the IAEA's Technical Working Group on Fast Reactors (TWGFR) and carried out from 1996 to 1999. One of the primary requirements for nuclear power plants and facilities is to ensure safety and the absence of damage under strong external dynamic loading from, for example, earthquakes. The designs of liquid metal cooled fast reactors (LMFRs) include systems which operate at low pressure and include components which are thin-walled and flexible. These systems and components could be considerably affected by earthquakes in seismic zones. Therefore, the IAEA through its advanced reactor technology development programme supports the activities of Member States to apply seismic isolation technology to LMFRs. The application of this technology to LMFRs and other nuclear plants and related facilities would offer the advantage that standard designs may be safely used in areas with a seismic risk. The technology may also provide a means of seismically upgrading nuclear facilities. Design analyses applied to such critical structures need to be firmly established, and the CRP provided a valuable tool in assessing their reliability. Ten organizations from India, Italy, Japan, the Republic of Korea, the Russian Federation, the United Kingdom, the United States of America and the European Commission co-operated in this CRP. This report documents the CRP activities, provides the main results and recommendations and includes the work carried out by the research groups at the participating institutes within the CRP on verification of their analysis methods for predicting the behaviour of seismically isolated nuclear structures

MR image-guided portal verification for brain treatment field

International Nuclear Information System (INIS)

Yin Fangfang; Gao Qinghuai; Xie Huchen; Nelson, Diana F.; Yu Yan; Kwok, W. Edmund; Totterman, Saara; Schell, Michael C.; Rubin, Philip

1998-01-01

Purpose: To investigate a method for the generation of digitally reconstructed radiographs directly from MR images (DRR-MRI) to guide a computerized portal verification procedure. Methods and Materials: Several major steps were developed to perform an MR image-guided portal verification procedure. Initially, a wavelet-based multiresolution adaptive thresholding method was used to segment the skin slice-by-slice in MR brain axial images. Some selected anatomical structures, such as target volume and critical organs, were then manually identified and were reassigned to relatively higher intensities. Interslice information was interpolated with a directional method to achieve comparable display resolution in three dimensions. Next, a ray-tracing method was used to generate a DRR-MRI image at the planned treatment position, and the ray tracing was simply performed on summation of voxels along the ray. The skin and its relative positions were also projected to the DRR-MRI and were used to guide the search of similar features in the portal image. A Canny edge detector was used to enhance the brain contour in both portal and simulation images. The skin in the brain portal image was then extracted using a knowledge-based searching technique. Finally, a Chamfer matching technique was used to correlate features between DRR-MRI and portal image. Results: The MR image-guided portal verification method was evaluated using a brain phantom case and a clinical patient case. Both DRR-CT and DRR-MRI were generated using CT and MR phantom images with the same beam orientation and then compared. The matching result indicated that the maximum deviation of internal structures was less than 1 mm. The segmented results for brain MR slice images indicated that a wavelet-based image segmentation technique provided a reasonable estimation for the brain skin. For the clinical patient case with a given portal field, the MR image-guided verification method provided an excellent match between

Status of safety issues at licensed power plants

International Nuclear Information System (INIS)

1991-03-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG series report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. The data contained in this report are a product of the NRC's Safety Issues Management System database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by personnel in the NRC regions. This report has been prepared in order to provide a comprehensive description of the implementation and verification status of all the TMI Action Plan requirements at licensed reactors, and to make this information available to other interested parties, including the public. A corollary purpose of this report is for it to serve as a follow-on to NUREG-0933, ''A Prioritization of Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed facilities

Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

2015-05-15

IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea.

Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

International Nuclear Information System (INIS)

Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun

2015-01-01

IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea

RELAP-7 Software Verification and Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Choi, Yong-Joon [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support; Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk, Reliability, and Regulatory Support

2014-09-25

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

Darcy Tools version 3.4. Verification, validation and demonstration

International Nuclear Information System (INIS)

Svensson, Urban

2010-12-01

DarcyTools is a computer code for simulation of flow and transport in porous and/or fractured media. The fractured media in mind is a fractured rock and the porous media the soil cover on the top of the rock; it is hence groundwater flows, which is the class of flows in mind. A number of novel methods and features form the present version of DarcyTools. In the verification studies, these methods are evaluated by comparisons with analytical solutions for idealized situations. The five verification groups (see Table 3-1 below), thus reflect the scope of DarcyTools. The present report will focus on the Verification, Validation and Demonstration of DarcyTools. Two accompanying reports cover other aspects: - Concepts, Methods and Equations, /Svensson et al. 2010/ (Hereafter denoted Report 1). - User's Guide, /Svensson and Ferry 2010/ (Hereafter denoted Report 2)

Darcy Tools version 3.4. Verification, validation and demonstration

Energy Technology Data Exchange (ETDEWEB)

Svensson, Urban (Computer-aided Fluid Engineering AB, Lyckeby (Sweden))

2010-12-15

DarcyTools is a computer code for simulation of flow and transport in porous and/or fractured media. The fractured media in mind is a fractured rock and the porous media the soil cover on the top of the rock; it is hence groundwater flows, which is the class of flows in mind. A number of novel methods and features form the present version of DarcyTools. In the verification studies, these methods are evaluated by comparisons with analytical solutions for idealized situations. The five verification groups (see Table 3-1 below), thus reflect the scope of DarcyTools. The present report will focus on the Verification, Validation and Demonstration of DarcyTools. Two accompanying reports cover other aspects: - Concepts, Methods and Equations, /Svensson et al. 2010/ (Hereafter denoted Report 1). - User's Guide, /Svensson and Ferry 2010/ (Hereafter denoted Report 2)

METHODS OF CONTROL DIPHTHERIA VACCINE SAFETY

Directory of Open Access Journals (Sweden)

Isayenko Ye. Yu

2016-12-01

Full Text Available Vaccination success depends not only on the timely coverage of threatened contingents, but also on the quality of vaccines. Every day, the requirements for security guarantees vaccines and their use guarantees of security increases. For the fast, reliable and independent scientific assessment of vaccine safety issues, WHO in 1999 created the Global Advisory Committee on Vaccine Safety. To enhance the capacity of pharmaceutical supervision in relation to vaccines in 2012 it was developed the Global Vaccine Safety Initiative. The main directions of the Global Vaccine Safety programs are considered in this review. It’s noted more strict requirements of Ukrainian pharmaceutical industry to produce public immunization drugs regulated Supplements to the State Pharmacopoeia of Ukraine, in comparison with other countries. This review considered diphtheria vaccine safety monitoring in the process of production according to the recommendations of the World Health Organization (WHO, described a subcutaneous method for determining the specific toxicity of the combined purified toxoid, characterized an intracutaneous method of determining of the presence of diphtheria toxin in each sample of the combined purified toxoid, that additionally used by some manufacturers. The definition of diphtheria toxin in dilutions of purified toxoid is presented. This review considered diphtheria vaccine safety monitoring in the process of production according to the recommendations of the World Health Organization (WHO, described a subcutaneous method for determining the specific toxicity of the combined purified toxoid, characterized an intracutaneous method of determining of the presence of diphtheria toxin in each sample of the combined purified toxoid, that additionally used by some manufacturers. The definition of diphtheria toxin in dilutions of purified toxoid is presented. As methods for determination of diphtheria toxin must be able to detect even a small amount

Verification of structural analysis computer codes in nuclear engineering

International Nuclear Information System (INIS)

Zebeljan, Dj.; Cizelj, L.

1990-01-01

Sources of potential errors, which can take place during use of finite element method based computer programs, are described in the paper. The magnitude of errors was defined as acceptance criteria for those programs. Error sources are described as they are treated by 'National Agency for Finite Element Methods and Standards (NAFEMS)'. Specific verification examples are used from literature of Nuclear Regulatory Commission (NRC). Example of verification is made on PAFEC-FE computer code for seismic response analyses of piping systems by response spectrum method. (author)

Verification and validation guidelines for high integrity systems. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Hecht, H.; Hecht, M.; Dinsmore, G.; Hecht, S.; Tang, D. [SoHaR, Inc., Beverly Hills, CA (United States)

1995-03-01

High integrity systems include all protective (safety and mitigation) systems for nuclear power plants, and also systems for which comparable reliability requirements exist in other fields, such as in the process industries, in air traffic control, and in patient monitoring and other medical systems. Verification aims at determining that each stage in the software development completely and correctly implements requirements that were established in a preceding phase, while validation determines that the overall performance of a computer system completely and correctly meets system requirements. Volume I of the report reviews existing classifications for high integrity systems and for the types of errors that may be encountered, and makes recommendations for verification and validation procedures, based on assumptions about the environment in which these procedures will be conducted. The final chapter of Volume I deals with a framework for standards in this field. Volume II contains appendices dealing with specific methodologies for system classification, for dependability evaluation, and for two software tools that can automate otherwise very labor intensive verification and validation activities.

Expert system verification and validation for nuclear power industry applications

International Nuclear Information System (INIS)

Naser, J.A.

1990-01-01

The potential for the use of expert systems in the nuclear power industry is widely recognized. The benefits of such systems include consistency of reasoning during off-normal situations when humans are under great stress, the reduction of times required to perform certain functions, the prevention of equipment failures through predictive diagnostics, and the retention of human expertise in performing specialized functions. The increased use of expert systems brings with it concerns about their reliability. Difficulties arising from software problems can affect plant safety, reliability, and availability. A joint project between EPRI and the US Nuclear Regulatory Commission is being initiated to develop a methodology for verification and validation of expert systems for nuclear power applications. This methodology will be tested on existing and developing expert systems. This effort will explore the applicability of conventional verification and validation methodologies to expert systems. The major area of concern will be certification of the knowledge base. This is expected to require new types of verification and validation techniques. A methodology for developing validation scenarios will also be studied

Verification and validation guidelines for high integrity systems. Volume 1

International Nuclear Information System (INIS)

Hecht, H.; Hecht, M.; Dinsmore, G.; Hecht, S.; Tang, D.

1995-03-01

High integrity systems include all protective (safety and mitigation) systems for nuclear power plants, and also systems for which comparable reliability requirements exist in other fields, such as in the process industries, in air traffic control, and in patient monitoring and other medical systems. Verification aims at determining that each stage in the software development completely and correctly implements requirements that were established in a preceding phase, while validation determines that the overall performance of a computer system completely and correctly meets system requirements. Volume I of the report reviews existing classifications for high integrity systems and for the types of errors that may be encountered, and makes recommendations for verification and validation procedures, based on assumptions about the environment in which these procedures will be conducted. The final chapter of Volume I deals with a framework for standards in this field. Volume II contains appendices dealing with specific methodologies for system classification, for dependability evaluation, and for two software tools that can automate otherwise very labor intensive verification and validation activities

49 CFR Appendix to Subpart H of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Non-North America-Domiciled...

Science.gov (United States)

2010-10-01

... safety audit will include: (1) Verification of available performance data and safety management programs; (2) Verification of a controlled substances and alcohol testing program consistent with part 40 of... Regulations, parts 382 through 399 of this subchapter, and the Federal Hazardous Material Regulations, parts...

Investigation of novel spent fuel verification system for safeguard application

Energy Technology Data Exchange (ETDEWEB)

Lee, Haneol; Yim, Man-Sung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Radioactive waste, especially spent fuel, is generated from the operation of nuclear power plants. The final stage of radioactive waste management is disposal which isolates radioactive waste from the accessible environment and allows it to decay. The safety, security, and safeguard of a spent fuel repository have to be evaluated before its operation. Many researchers have evaluated the safety of a repository. These researchers calculated dose to public after the repository is closed depending on their scenario. Because most spent fuel repositories are non-retrievable, research on security or safeguards of spent fuel repositories have to be performed. Design based security or safeguard have to be developed for future repository designs. This study summarizes the requirements of future spent fuel repositories especially safeguards, and suggests a novel system which meets the safeguard requirements. Applying safeguards to a spent fuel repository is becoming increasingly important. The future requirements for a spent fuel repository are suggested by several expert groups, such as ASTOR in IAEA. The requirements emphasizes surveillance and verification. The surveillance and verification of spent fuel is currently accomplished by using the Cerenkov radiation detector while spent fuel is being stored in a fuel pool. This research investigated an advanced spent fuel verification system using a system which converts spent fuel radiation into electricity. The system generates electricity while it is conveyed from a transportation cask to a disposal cask. The electricity conversion system was verified in a lab scale experiment using an 8.51GBq Cs-137 gamma source.

Investigation of novel spent fuel verification system for safeguard application

International Nuclear Information System (INIS)

Lee, Haneol; Yim, Man-Sung

2016-01-01

Radioactive waste, especially spent fuel, is generated from the operation of nuclear power plants. The final stage of radioactive waste management is disposal which isolates radioactive waste from the accessible environment and allows it to decay. The safety, security, and safeguard of a spent fuel repository have to be evaluated before its operation. Many researchers have evaluated the safety of a repository. These researchers calculated dose to public after the repository is closed depending on their scenario. Because most spent fuel repositories are non-retrievable, research on security or safeguards of spent fuel repositories have to be performed. Design based security or safeguard have to be developed for future repository designs. This study summarizes the requirements of future spent fuel repositories especially safeguards, and suggests a novel system which meets the safeguard requirements. Applying safeguards to a spent fuel repository is becoming increasingly important. The future requirements for a spent fuel repository are suggested by several expert groups, such as ASTOR in IAEA. The requirements emphasizes surveillance and verification. The surveillance and verification of spent fuel is currently accomplished by using the Cerenkov radiation detector while spent fuel is being stored in a fuel pool. This research investigated an advanced spent fuel verification system using a system which converts spent fuel radiation into electricity. The system generates electricity while it is conveyed from a transportation cask to a disposal cask. The electricity conversion system was verified in a lab scale experiment using an 8.51GBq Cs-137 gamma source

Safety and Waste Management for SAM Pathogen Methods

Science.gov (United States)

The General Safety and Waste Management page offers section-specific safety and waste management details for the pathogens included in EPA's Selected Analytical Methods for Environmental Remediation and Recovery (SAM).

Safety and Waste Management for SAM Biotoxin Methods

Science.gov (United States)

The General Safety and Waste Management page offers section-specific safety and waste management details for the biotoxins included in EPA's Selected Analytical Methods for Environmental Remediation and Recovery (SAM).

Challenges for effective WMD verification

International Nuclear Information System (INIS)

Andemicael, B.

2006-01-01

already awash in fissile material and is increasingly threatened by the possible consequences of illicit trafficking in such material. The chemical field poses fewer problems. The ban on chemical weapons is a virtually complete post-Cold War regime, with state-of-the-art concepts and procedures of verification resulting from decades of negotiation. The detection of prohibited materials and activities is the common goal of the nuclear and chemical regimes for which the most intrusive and intensive procedures are activated by the three organizations. Accounting for the strictly peaceful application of dual-use items constitutes the bulk of the work of the inspectorates at the IAEA and the OPCW. A common challenge in both fields is the advance of science and technology in the vast nuclear and chemical industries and the ingenuity of some determined proliferators to deceive by concealing illicit activities under legitimate ones. Inspection procedures and technologies need to keep up with the requirement for flexibility and adaptation to change. The common objective of the three organizations is to assemble and analyze all relevant information in order to conclude reliably whether a State is or is not complying with its treaty obligations. The positive lessons learned from the IAEA's verification experience today are valuable in advancing concepts and technologies that might also benefit the other areas of WMD verification. Together with the emerging, more comprehensive verification practice of the OPCW, they may provide a useful basis for developing common standards, which may in turn help in evaluating the cost-effectiveness of verification methods for the Biological and Toxin Weapons Convention and other components of a WMD control regime

Safety and Waste Management for SAM Chemistry Methods

Science.gov (United States)

The General Safety and Waste Management page offers section-specific safety and waste management details for the chemical analytes included in EPA's Selected Analytical Methods for Environmental Remediation and Recovery (SAM).

Safety and Waste Management for SAM Radiochemical Methods

Science.gov (United States)

The General Safety and Waste Management page offers section-specific safety and waste management details for the radiochemical analytes included in EPA's Selected Analytical Methods for Environmental Remediation and Recovery (SAM).

Software verification for nuclear industry

International Nuclear Information System (INIS)

Wilburn, N.P.

1985-08-01

Why verification of software products throughout the software life cycle is necessary is considered. Concepts of verification, software verification planning, and some verification methodologies for products generated throughout the software life cycle are then discussed

The design of verification regimes

International Nuclear Information System (INIS)

Gallagher, N.W.

1991-01-01

Verification of a nuclear agreement requires more than knowledge of relevant technologies and institutional arrangements. It also demands thorough understanding of the nature of verification and the politics of verification design. Arms control efforts have been stymied in the past because key players agreed to verification in principle, only to disagree radically over verification in practice. In this chapter, it is shown that the success and stability of arms control endeavors can be undermined by verification designs which promote unilateral rather than cooperative approaches to security, and which may reduce, rather than enhance, the security of both sides. Drawing on logical analysis and practical lessons from previous superpower verification experience, this chapter summarizes the logic and politics of verification and suggests implications for South Asia. The discussion begins by determining what properties all forms of verification have in common, regardless of the participants or the substance and form of their agreement. Viewing verification as the political process of making decisions regarding the occurrence of cooperation points to four critical components: (1) determination of principles, (2) information gathering, (3) analysis and (4) projection. It is shown that verification arrangements differ primarily in regards to how effectively and by whom these four stages are carried out

Knowledge base verification based on enhanced colored petri net

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Hyun; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

Verification is a process aimed at demonstrating whether a system meets it`s specified requirements. As expert systems are used in various applications, the knowledge base verification of systems takes an important position. The conventional Petri net approach that has been studied recently in order to verify the knowledge base is found that it is inadequate to verify the knowledge base of large and complex system, such as alarm processing system of nuclear power plant. Thus, we propose an improved method that models the knowledge base as enhanced colored Petri net. In this study, we analyze the reachability and the error characteristics of the knowledge base and apply the method to verification of simple knowledge base. 8 refs., 4 figs. (Author)

Knowledge base verification based on enhanced colored petri net

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Hyun; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

Verification is a process aimed at demonstrating whether a system meets it`s specified requirements. As expert systems are used in various applications, the knowledge base verification of systems takes an important position. The conventional Petri net approach that has been studied recently in order to verify the knowledge base is found that it is inadequate to verify the knowledge base of large and complex system, such as alarm processing system of nuclear power plant. Thus, we propose an improved method that models the knowledge base as enhanced colored Petri net. In this study, we analyze the reachability and the error characteristics of the knowledge base and apply the method to verification of simple knowledge base. 8 refs., 4 figs. (Author)

Verification of Triple Modular Redundancy (TMR) Insertion for Reliable and Trusted Systems

Science.gov (United States)

Berg, Melanie; LaBel, Kenneth A.

2016-01-01

We propose a method for TMR insertion verification that satisfies the process for reliable and trusted systems. If a system is expected to be protected using TMR, improper insertion can jeopardize the reliability and security of the system. Due to the complexity of the verification process, there are currently no available techniques that can provide complete and reliable confirmation of TMR insertion. This manuscript addresses the challenge of confirming that TMR has been inserted without corruption of functionality and with correct application of the expected TMR topology. The proposed verification method combines the usage of existing formal analysis tools with a novel search-detect-and-verify tool. Field programmable gate array (FPGA),Triple Modular Redundancy (TMR),Verification, Trust, Reliability,

Applying formal method to design of nuclear power plant embedded protection system

International Nuclear Information System (INIS)

Kim, Jin Hyun; Kim, Il Gon; Sung, Chang Hoon; Choi, Jin Young; Lee, Na Young

2001-01-01

Nuclear power embedded protection systems is a typical safety-critical system, which detects its failure and shutdowns its operation of nuclear reactor. These systems are very dangerous so that it absolutely requires safety and reliability. Therefore nuclear power embedded protection system should fulfill verification and validation completely from the design stage. To develop embedded system, various V and V method have been provided and especially its design using Formal Method is studied in other advanced country. In this paper, we introduce design method of nuclear power embedded protection systems using various Formal-Method in various respect following nuclear power plant software development guideline

On Backward-Style Anonymity Verification

Science.gov (United States)

Kawabe, Yoshinobu; Mano, Ken; Sakurada, Hideki; Tsukada, Yasuyuki

Many Internet services and protocols should guarantee anonymity; for example, an electronic voting system should guarantee to prevent the disclosure of who voted for which candidate. To prove trace anonymity, which is an extension of the formulation of anonymity by Schneider and Sidiropoulos, this paper presents an inductive method based on backward anonymous simulations. We show that the existence of an image-finite backward anonymous simulation implies trace anonymity. We also demonstrate the anonymity verification of an e-voting protocol (the FOO protocol) with our backward anonymous simulation technique. When proving the trace anonymity, this paper employs a computer-assisted verification tool based on a theorem prover.

Uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. Final report, February 16, 1990--December 31, 1994

International Nuclear Information System (INIS)

Busch, R.D.

1995-01-01

Dr. Robert Busch of the Department of Chemical and Nuclear Engineering was the principal investigator on this project with technical direction provided by the staff in the Nuclear Criticality Safety Group at Los Alamos. During the period of the contract, he had a number of graduate and undergraduate students working on subtasks. The objective of this work was to develop information on uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. During the first year of this project, most of the work was focused on setting up the SUN SPARC-1 Workstation and acquiring the literature which described the critical experiments. By august 1990, the Workstation was operational with the current version of TWODANT loaded on the system. MCNP, version 4 tape was made available from Los Alamos late in 1990. Various documents were acquired which provide the initial descriptions of the critical experiments under consideration as benchmarks. The next four years were spent working on various benchmark projects. A number of publications and presentations were made on this material. These are briefly discussed in this report

Teen worker safety training: methods used, lessons taught, and time spent.

Science.gov (United States)

Zierold, Kristina M

2015-05-01

Safety training is strongly endorsed as one way to prevent teens from performing dangerous tasks at work. The objective of this mixed methods study was to characterize the safety training that teenagers receive on the job. From 2010 through 2012, focus groups and a cross-sectional survey were conducted with working teens. The top methods of safety training reported were safety videos (42 percent) and safety lectures (25 percent). The top lessons reported by teens were "how to do my job" and "ways to spot hazards." Males, who were more likely to do dangerous tasks, received less safety training than females. Although most teens are getting safety training, it is inadequate. Lessons addressing safety behaviors are missing, training methods used are minimal, and the time spent is insignificant. More research is needed to understand what training methods and lessons should be used, and the appropriate safety training length for effectively preventing injury in working teens. In addition, more research evaluating the impact of high-quality safety training compared to poor safety training is needed to determine the best training programs for teens. © The Author(s) 2015 Reprints and permissions: sagepub.co.uk/journalsPermissions.nav.

Physics Verification Overview

Energy Technology Data Exchange (ETDEWEB)

Doebling, Scott William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-12

The purpose of the verification project is to establish, through rigorous convergence analysis, that each ASC computational physics code correctly implements a set of physics models and algorithms (code verification); Evaluate and analyze the uncertainties of code outputs associated with the choice of temporal and spatial discretization (solution or calculation verification); and Develop and maintain the capability to expand and update these analyses on demand. This presentation describes project milestones.

Verification and Validation of RADTRAN 5.5.

Energy Technology Data Exchange (ETDEWEB)

Osborn, Douglas.; Weiner, Ruth F.; Mills, George Scott; Hamp, Steve C.

2005-02-01

This document contains a description of the verification and validation process used for the RADTRAN 5.5 code. The verification and validation process ensured the proper calculational models and mathematical and numerical methods were used in the RADTRAN 5.5 code for the determination of risk and consequence assessments. The differences between RADTRAN 5 and RADTRAN 5.5 are the addition of tables, an expanded isotope library, and the additional User-Defined meteorological option for accident dispersion. 3

Formal Development and Verification of a Distributed Railway Control System

DEFF Research Database (Denmark)

Haxthausen, Anne Elisabeth; Peleska, Jan

1999-01-01

In this article we introduce the concept for a distributed railway control system and present the specification and verification of the main algorithm used for safe distributed control. Our design and verification approach is based on the RAISE method, starting with highly abstract algebraic...

Formal Development and Verification of a Distributed Railway Control System

DEFF Research Database (Denmark)

Haxthausen, Anne Elisabeth; Peleska, Jan

1998-01-01

In this article we introduce the concept for a distributed railway control system and present the specification and verification of the main algorithm used for safe distributed control. Our design and verification approach is based on the RAISE method, starting with highly abstract algebraic spec...

Comparison, with regard to safety, between a hard-wired reactor protection system and a computerized protection system. Pt. 1

International Nuclear Information System (INIS)

Buettner, W.E.

1976-07-01

The study compares a conventional hard-wired dynamic reactor protection system with a computerized protection system. In the comparison, only the unequivocally safety-oriented protection actions are considered. In the first part, the different structures of both systems and the method of verification for their functional safety will be described. In the second part, the mean unavailability in case of demand for both systems under defined conditions will be determined. (orig.) [de

Analysis of tank safety with propane-butane on LPG distribution station

Directory of Open Access Journals (Sweden)

Krzysiak Zbigniew

2017-12-01

Full Text Available An analysis of the risk of failure in the safety valve – tank with propane-butane (LPG system has been conducted. An uncontrolled outflow of liquid LPG, caused by a failure of the above mentioned system has been considered as a threat. The main research goal of the study is the hazardous analysis of propane-butane gas outflow for the safety valve – LPG tank system. The additional goal is the development of an useful method to fast identify the hazard of a mismatched safety valve. The results of the research analysis have confirmed that safety valves are basic protection of the installation (tank against failures that can lead to loss of life, material damage and further undesired costs of their unreliability. That is why a new, professional computer program has been created that allows for the selection of safety valves or for the verification of a safety valve selection in installations where any technical or technological changes have been made.

An evaluation of the management system verification pilot at Hanford

International Nuclear Information System (INIS)

Briggs, C.R.; Ramonas, L.; Westendorf, W.

1998-01-01

The Chemical Management System (CMS), currently under development at Hanford, was used as the ''test program'' for pilot testing the value added aspects of the Chemical Manufacturers Association's (CMA) Management Systems Verification (MSV) process. The MSV process, which was developed by CMA's member chemical companies specifically as a tool to assist in the continuous improvement of environment, safety and health (ESH) performance, represents a commercial sector ''best practice'' for evaluating ESH management systems. The primary purpose of Hanford's MSV Pilot was to evaluate the applicability and utility of the MSV process in the Department of Energy (DOE) environment. However, because the Integrated Safety Management System (ISMS) is the framework for ESH management at Hanford and at all DOE sites, the pilot specifically considered the MSV process in the context of a possible future adjunct to Integrated Safety Management System Verification (ISMSV) efforts at Hanford and elsewhere within the DOE complex. The pilot involved the conduct of two-hour interviews with four separate panels of individuals with functional responsibilities related to the CMS including the Department of Energy Richland Operations (DOE-RL), Fluor Daniel Hanford (FDH) and FDH's major subcontractors (MSCS). A semi-structured interview process was employed by the team of three ''verifiers'' who directed open-ended questions to the panels regarding the development, integration and effectiveness of management systems necessary to ensure the sustainability of the CMS effort. An ''MSV Pilot Effectiveness Survey'' also was completed by each panel participant immediately following the interview

The Demon-Angel method in systematic safety assessment

International Nuclear Information System (INIS)

Vassakis, A.G.

1999-01-01

Since 'design for safety' of large engineering systems with a high level of innovation requires an exhaustive safety analysis and since a subsequent corrective action may become a very large bottleneck in improving such a system, fundamental knowledge in designing safe systems is essential. The lack of any 'rule of thumb' makes such knowledge a matter of personal experience rather than the subject of an academic course. This paper proposes a new method for the theoretical safety study of different system configurations independently of any particular application. This method aims to help the construction of a 'rule of thumb' for what is a safe system and what is not. The Demon and Angel ideas are explained and the schematic presentation of these elements is introduced. Four representative case studies demonstrate the use of this method

Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

Energy Technology Data Exchange (ETDEWEB)

Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P. [and others

1996-12-01

This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising.

Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

International Nuclear Information System (INIS)

Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P.

1996-01-01

This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising

Improvement and verification of fast-reactor safety-analysis techniques. Final report

International Nuclear Information System (INIS)

Barker, D.H.

1981-12-01

The work involved on this project took place between March 1, 1975 and December 31, 1981. The work resulted in two PhD and one Masters Theses. Part I was the Verification and Applicability Studies for the VENUS-II LMFBR Disassembly Code. These tests showed that the VENUS-II code closely predicted the energy release in all three tests chosen for analysis. Part II involved the chemical simulation of pool dispersion in the transition phase of an HCDA. Part III involved the reaction of an internally heated fluid and the vessel walls

Balance between qualitative and quantitative verification methods

International Nuclear Information System (INIS)

Nidaira, Kazuo

2012-01-01

The amount of inspection effort for verification of declared nuclear material needs to be optimized in the situation where qualitative and quantitative measures are applied. Game theory was referred to investigate the relation of detection probability and deterrence of diversion. Payoffs used in the theory were quantified for cases of conventional safeguards and integrated safeguards by using AHP, Analytical Hierarchy Process. Then, it became possible to estimate detection probability under integrated safeguards which had equivalent deterrence capability for detection probability under conventional safeguards. In addition the distribution of inspection effort for qualitative and quantitative measures was estimated. Although the AHP has some ambiguities in quantifying qualitative factors, its application to optimization in safeguards is useful to reconsider the detection probabilities under integrated safeguards. (author)

A document-driven method for certifying scientific computing software for use in nuclear safety analysis

International Nuclear Information System (INIS)

Smith, W. Spencer; Koothoor, Mimitha

2016-01-01

This paper presents a documentation and development method to facilitate the certification of scientific computing software used in the safety analysis of nuclear facilities. To study the problems faced during quality assurance and certification activities, a case study was performed on legacy software used for thermal analysis of a fuel pin in a nuclear reactor. Although no errors were uncovered in the code, 27 issues of incompleteness and inconsistency were found with the documentation. This work proposes that software documentation follow a rational process, which includes a software requirements specification following a template that is reusable, maintainable, and understandable. To develop the design and implementation, this paper suggests literate programming as an alternative to traditional structured programming. Literate programming allows for documenting of numerical algorithms and code together in what is termed the literate programmer's manual. This manual is developed with explicit traceability to the software requirements specification. The traceability between the theory, numerical algorithms, and implementation facilitates achieving completeness and consistency, as well as simplifies the process of verification and the associated certification

A document-driven method for certifying scientific computing software for use in nuclear safety analysis

Energy Technology Data Exchange (ETDEWEB)

Smith, W. Spencer; Koothoor, Mimitha [Computing and Software Department, McMaster University, Hamilton (Canada)

2016-04-15

This paper presents a documentation and development method to facilitate the certification of scientific computing software used in the safety analysis of nuclear facilities. To study the problems faced during quality assurance and certification activities, a case study was performed on legacy software used for thermal analysis of a fuel pin in a nuclear reactor. Although no errors were uncovered in the code, 27 issues of incompleteness and inconsistency were found with the documentation. This work proposes that software documentation follow a rational process, which includes a software requirements specification following a template that is reusable, maintainable, and understandable. To develop the design and implementation, this paper suggests literate programming as an alternative to traditional structured programming. Literate programming allows for documenting of numerical algorithms and code together in what is termed the literate programmer's manual. This manual is developed with explicit traceability to the software requirements specification. The traceability between the theory, numerical algorithms, and implementation facilitates achieving completeness and consistency, as well as simplifies the process of verification and the associated certification.

Verification Failures: What to Do When Things Go Wrong

Science.gov (United States)

Bertacco, Valeria

Every integrated circuit is released with latent bugs. The damage and risk implied by an escaped bug ranges from almost imperceptible to potential tragedy; unfortunately it is impossible to discern within this range before a bug has been exposed and analyzed. While the past few decades have witnessed significant efforts to improve verification methodology for hardware systems, these efforts have been far outstripped by the massive complexity of modern digital designs, leading to product releases for which an always smaller fraction of system's states has been verified. The news of escaped bugs in large market designs and/or safety critical domains is alarming because of safety and cost implications (due to replacements, lawsuits, etc.).

Using SysML for verification and validation planning on the Large Synoptic Survey Telescope (LSST)

Science.gov (United States)

Selvy, Brian M.; Claver, Charles; Angeli, George

2014-08-01

This paper provides an overview of the tool, language, and methodology used for Verification and Validation Planning on the Large Synoptic Survey Telescope (LSST) Project. LSST has implemented a Model Based Systems Engineering (MBSE) approach as a means of defining all systems engineering planning and definition activities that have historically been captured in paper documents. Specifically, LSST has adopted the Systems Modeling Language (SysML) standard and is utilizing a software tool called Enterprise Architect, developed by Sparx Systems. Much of the historical use of SysML has focused on the early phases of the project life cycle. Our approach is to extend the advantages of MBSE into later stages of the construction project. This paper details the methodology employed to use the tool to document the verification planning phases, including the extension of the language to accommodate the project's needs. The process includes defining the Verification Plan for each requirement, which in turn consists of a Verification Requirement, Success Criteria, Verification Method(s), Verification Level, and Verification Owner. Each Verification Method for each Requirement is defined as a Verification Activity and mapped into Verification Events, which are collections of activities that can be executed concurrently in an efficient and complementary way. Verification Event dependency and sequences are modeled using Activity Diagrams. The methodology employed also ties in to the Project Management Control System (PMCS), which utilizes Primavera P6 software, mapping each Verification Activity as a step in a planned activity. This approach leads to full traceability from initial Requirement to scheduled, costed, and resource loaded PMCS task-based activities, ensuring all requirements will be verified.

Experimental study on design verification of new concept for integral reactor safety system

International Nuclear Information System (INIS)

Chung, Moon Ki; Choi, Ki Yong; Park, Hyun Sik; Cho, Seok; Park, Choon Kyung; Lee, Sung Jae; Song, Chul Hwa

2004-01-01

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the Steam Generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant

FLIGHT SAFETY MANAGEMENT PROBLEMS AND EVALUATION OF FLIGHT SAFETY LEVEL OF AN AVIATION ENTERPRISE

Directory of Open Access Journals (Sweden)

B. V. Zubkov

2017-01-01

Full Text Available This article is devoted to studying the problem of safety management system (SMS and evaluating safety level of an aviation enterprise.This article discusses the problems of SMS, presented at the 41st meeting of the Russian Aviation Production Commanders Club in June 2014 in St. Petersburg in connection with the verification of the status of the CA of the Russian Federation by the International Civil Aviation Organization (ICAO in the same year, a set of urgent measures to eliminate the deficiencies identified in the current safety management system by participants of this meeting were proposed.In addition, the problems of evaluating flight safety level based on operation data of an aviation enterprise were analyzed. This analysis made it possible to take into account the problems listed in this article as a tool for a comprehensive study of SMS parameters and allows to analyze the quantitative indicators of the flights safety level.The concepts of Acceptable Safety Level (ASL indicators are interpreted differently depending on the available/applicable methods of their evaluation and how to implement them in SMS. However, the indicators for assessing ASL under operational condition at the aviation enterprise should become universal. Currently, defined safety levels and safety indicators are not yet established functionally and often with distorted underrepresented models describing their contextual contents, as well as ways of integrating them into SMS aviation enterprise.The results obtained can be used for better implementation of SMS and solving problems determining the aviation enterprise technical level of flight safety.

Software Safety Life cycle and Method of POSAFE-Q System

International Nuclear Information System (INIS)

Lee, Jang-Soo; Kwon, Kee-Choon

2006-01-01

This paper describes the relationship between the overall safety life cycle and the software safety life cycle during the development of the software based safety systems of Nuclear Power Plants. This includes the design and evaluation activities of components as well as the system. The paper also compares the safety life cycle and planning activities defined in IEC 61508 with those in IEC 60880, IEEE 7-4.3.2, and IEEE 1228. Using the KNICS project as an example, software safety life cycle and safety analysis methods applied to the POSAFE-Q are demonstrated. KNICS software safety life cycle is described by comparing to the software development, testing, and safety analysis process with international standards. The safety assessment of the software for POSAFE-Q is a joint Korean German project. The assessment methods applied in the project and the experiences gained from this project are presented