Probabilistic methods of optimization of scheduled tests for heat equipment of safety systems of reactor at full power

International Nuclear Information System (INIS)

Bilej, D.V.; Fridman, N.A.; Kolykhanov, V.N.; Skalozubov, V.I.

2004-01-01

This article generalises the basic results of a long-term teamwork with respect to a scientific and technical substantiation of perfection of the regulations of safe operation power units with VVER. This perfection is concerning a periodicity and volumes of tests of safety systems when a reactor works at full power. The article shows that the application of the probabilistic approaches connected to minimisation of a risk criterion function is an effective methodical base for the optimisation. For certain safety systems of serial power units with VVER 1000 the results of calculated substantiations are presented

Determination of the number of software tests using probabilistic safety assessment

International Nuclear Information System (INIS)

Kang, H. K.; Seong, T. Y.; Lee, K. Y.

2000-01-01

The broader usage of digital equipment in nuclear power plants gives rise to the safety problems of software. The field test should be performed before the software is used in critical applications because it is well known that software shows non-linear response when it is applied to different target systems in different environment. In the case of safety-critical applications, the result of tests contains usually zero failure case and the satisfiable number of tests is hard to be determined. In this paper, we suggests the method to determine the number of software tests without failure using the probabilistic safety assessment. From the result of the probabilistic safety assessment on total system, the desirable unavailability of software is calculated and the number of tests is determined

Quantification of Safety-Critical Software Test Uncertainty

International Nuclear Information System (INIS)

Khalaquzzaman, M.; Cho, Jaehyun; Lee, Seung Jun; Jung, Wondea

2015-01-01

The method, conservatively assumes that the failure probability of a software for the untested inputs is 1, and the failure probability turns in 0 for successful testing of all test cases. However, in reality the chance of failure exists due to the test uncertainty. Some studies have been carried out to identify the test attributes that affect the test quality. Cao discussed the testing effort, testing coverage, and testing environment. Management of the test uncertainties was discussed in. In this study, the test uncertainty has been considered to estimate the software failure probability because the software testing process is considered to be inherently uncertain. A reliability estimation of software is very important for a probabilistic safety analysis of a digital safety critical system of NPPs. This study focused on the estimation of the probability of a software failure that considers the uncertainty in software testing. In our study, BBN has been employed as an example model for software test uncertainty quantification. Although it can be argued that the direct expert elicitation of test uncertainty is much simpler than BBN estimation, however the BBN approach provides more insights and a basis for uncertainty estimation

9 CFR 113.39 - Cat safety tests.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Cat safety tests. 113.39 Section 113... Procedures § 113.39 Cat safety tests. The safety tests provided in this section shall be conducted when... recommended for use in cats. (a) The cat safety test provided in this paragraph shall be used when the Master...

Standard test method for dynamic tear testing of metallic materials

CERN Document Server

American Society for Testing and Materials. Philadelphia

1983-01-01

1.1 This test method covers the dynamic tear (DT) test using specimens that are 3/16 in. to 5/8 in. (5 mm to 16 mm) inclusive in thickness. 1.2 This test method is applicable to materials with a minimum thickness of 3/16 in. (5 mm). 1.3 The pressed-knife procedure described for sharpening the notch tip generally limits this test method to materials with a hardness level less than 36 HRC. Note 1—The designation 36 HRC is a Rockwell hardness number of 36 on Rockwell C scale as defined in Test Methods E 18. 1.4 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Safety tests carried out at Cadarache. Sodium fires

International Nuclear Information System (INIS)

Fruchard, M.

1976-01-01

Safety test on sodium fires developed at the Cadarache Nuclear Centre by the Department of Nuclear Safety, section of safety experiments on radioactivity transfer are conducted in two main directions: analysis of the behavior and thermodynamic consequences of accidental fires, working on the basis of typical experimental results; research and development of methods and equipment to control and if possible extinguish these fires. The most important part of this programme is concerned with the sodium pool fires which would result from the failure of a secondary coolant circuit pipe [fr

Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2003-03-01

Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003. (author)

Safety of pulmonary function testing

DEFF Research Database (Denmark)

Roberts, Cara; Ward, Simon; Walsted, Emil

2017-01-01

BACKGROUND: Pulmonary function testing (PFT) is a key investigation in the evaluation of individuals with respiratory symptoms; however, the safety of routine and specialised PFT testing has not been reported in a large data set. Using patient safety incident (PSI) records, we aimed to assess risk...... was rated using the NHS National Patient Safety Agency and any hospital admission reported. RESULTS: There were 119 PSIs reported from 186 000 PFT; that is, 0.6 PSIs per 1000 tests. Cardiopulmonary PSIs were 3.3 times more likely to occur than non-cardiopulmonary (95% CI 2.17 to 5.12). Syncope was the most...

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)

Towards a Usability and Error "Safety Net": A Multi-Phased Multi-Method Approach to Ensuring System Usability and Safety.

Science.gov (United States)

Kushniruk, Andre; Senathirajah, Yalini; Borycki, Elizabeth

2017-01-01

The usability and safety of health information systems have become major issues in the design and implementation of useful healthcare IT. In this paper we describe a multi-phased multi-method approach to integrating usability engineering methods into system testing to ensure both usability and safety of healthcare IT upon widespread deployment. The approach involves usability testing followed by clinical simulation (conducted in-situ) and "near-live" recording of user interactions with systems. At key stages in this process, usability problems are identified and rectified forming a usability and technology-induced error "safety net" that catches different types of usability and safety problems prior to releasing systems widely in healthcare settings.

FOOD SAFETY TESTING LABORATORY

Data.gov (United States)

Federal Laboratory Consortium — This laboratory develops screening assays, tests and modifies biosensor equipment, and optimizes food safety testing protocols for the military and civilian sector...

Neurovirulence safety testing of mumps vaccines--historical perspective and current status.

Science.gov (United States)

Rubin, S A; Afzal, M A

2011-04-05

Many live, attenuated viral vaccines are derived from wild type viruses with known neurovirulent properties. To assure the absence of residual neurotoxicity, pre-clinical neurovirulence safety testing of candidate vaccines is performed. For mumps virus, a highly neurotropic virus, neurovirulence safety testing is performed in monkeys. However, laboratory studies suggest an inability of this test to correctly discern among virus strains of varying neurovirulence potential in man, and, further, some vaccines found to be neuroattenuated in monkeys were later found to be neurovirulent in humans when administered in large numbers. Over the past decade, concerted efforts have been made to replace monkey-based neurovirulence safety testing with more informative, alternative methods. This review summarizes the current status of mumps vaccine neurovirulence safety testing and insights into models currently approved and those under development. Published by Elsevier Ltd.

Valve testing for UK PWR safety applications

International Nuclear Information System (INIS)

George, P.T.; Bryant, S.

1989-01-01

Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

9 CFR 113.40 - Dog safety tests.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Dog safety tests. 113.40 Section 113... Procedures § 113.40 Dog safety tests. The safety tests provided in this section shall be conducted when... recommended for use in dogs. Serials which are not found to be satisfactory when tested pursuant to the...

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

Standard Test Method for Wet Insulation Integrity Testing of Photovoltaic Arrays

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This test method covers a procedure to determine the insulation resistance of a photovoltaic (PV) array (or its component strings), that is, the electrical resistance between the array's internal electrical components and is exposed, electrically conductive, non-current carrying parts and surfaces of the array. 1.2 This test method does not establish pass or fail levels. The determination of acceptable or unacceptable results is beyond the scope of this test method. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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.

9 CFR 113.33 - Mouse safety tests.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Mouse safety tests. 113.33 Section 113.33 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE... Procedures § 113.33 Mouse safety tests. One of the mouse safety tests provided in this section shall be...

A new method for testing and evaluating grassy airfields and its effects upon flying safety

Directory of Open Access Journals (Sweden)

Jarosław PYTKA

2017-06-01

Full Text Available The paper contains a review of methods and technologies developed during a research project entitled “Methods of testing and evaluation of grassy airfields”, conducted at the Lublin University of Technology between 2011 and 2014. Based on the terramechanical studies of wheel-soil interactions, the authors have developed a method for the determination of wheel-grass friction and rolling resistance coefficients, which are of critical importance for the ground performance of an airplane. Moreover, a mobile application has been created for use by pilots, controllers or airfield administration crew. The application connects online with a weather service to gather atmospheric data as inputs for a mathematical model that produces a real-time cone index (CI value for a given airfield. The paper also discusses the applicability of the method within the air transportation system, as well as possible effects of the described technology upon the safety of flight operations on grassy airfields.

Computational methods for nuclear criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.

1992-01-01

Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

Standard Test Method for Oxyacetylene Ablation Testing of Thermal Insulation Materials

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers the screening of ablative materials to determine the relative thermal insulation effectiveness when tested as a flat panel in an environment of a steady flow of hot gas provided by an oxyacetylene burner. 1.2 This test method should be used to measure and describe the properties of materials, products, or assemblies in response to heat and flame under controlled laboratory conditions and should not be used to describe or appraise the fire hazard of materials, products, or assemblies under actual fire conditions. However, results of this test method may be used as elements of a fire risk assessment which takes into account all of the factors which are pertinent to an assessment of the fire hazard of a particular end use. 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 appropriate safety and health practices and determine the applicability of regulatory limi...

Validation of innovative technologies and strategies for regulatory safety assessment methods: challenges and opportunities.

Science.gov (United States)

Stokes, William S; Wind, Marilyn

2010-01-01

Advances in science and innovative technologies are providing new opportunities to develop test methods and strategies that may improve safety assessments and reduce animal use for safety testing. These include high throughput screening and other approaches that can rapidly measure or predict various molecular, genetic, and cellular perturbations caused by test substances. Integrated testing and decision strategies that consider multiple types of information and data are also being developed. Prior to their use for regulatory decision-making, new methods and strategies must undergo appropriate validation studies to determine the extent that their use can provide equivalent or improved protection compared to existing methods and to determine the extent that reproducible results can be obtained in different laboratories. Comprehensive and optimal validation study designs are expected to expedite the validation and regulatory acceptance of new test methods and strategies that will support improved safety assessments and reduced animal use for regulatory testing.

Standard test method for galling resistance of material couples

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers a laboratory test that ranks the galling resistance of material couples using a quantitative measure. Bare metals, alloys, nonmetallic materials, coatings, and surface modified materials may be evaluated by this test method. 1.2 This test method is not designed for evaluating the galling resistance of material couples sliding under lubricated conditions, because galling usually will not occur under lubricated sliding conditions using this test method. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

International Nuclear Information System (INIS)

Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young; No, Young Hun; Song, Seung Hwan

2006-01-01

The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants

Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); No, Young Hun; Song, Seung Hwan [POSCON, Seoul (Korea, Republic of)

2006-07-01

The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants.

Test Bed for Safety Assessment of New e-Navigation Systems

Directory of Open Access Journals (Sweden)

Axel Hahn

2014-12-01

Full Text Available New e-navigation strains require new technologies, new infrastructures and new organizational structures on bridge, on shore as well as in the cloud. Suitable engineering and safety/risk assessment methods facilitate these efforts. Understanding maritime transportation as a sociotechnical system allows the application of system-engineering methods. Formal, simulation based and in situ verification and validation of e-navigation technologies are important methods to obtain system safety and reliability. The modelling and simulation toolset HAGGIS provides methods for system specification and formal risk analysis. It provides a modelling framework for processes, fault trees and generic hazard specification and a physical world and maritime traffic simulation system. HAGGIS is accompanied by the physical test bed LABSKAUS which implements a physical test bed. The test bed provides reference ports and waterways in combination with an experimental Vessel Traffic Services (VTS system and a mobile integrated bridge: This enables in situ experiments for technological evaluation, testing, ground research and demonstration. This paper describes an integrated seamless approach for developing new e-navigation technologies starting with simulation based assessment and ending in physical real world demonstrations

Standard test method for drop-weight tear tests of ferritic steels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This test method covers drop-weight tear tests (DWTT) on ferritic steels with thicknesses between 3.18 and 19.1 mm. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Usability Methods for Ensuring Health Information Technology Safety: Evidence-Based Approaches. Contribution of the IMIA Working Group Health Informatics for Patient Safety.

Science.gov (United States)

Borycki, E; Kushniruk, A; Nohr, C; Takeda, H; Kuwata, S; Carvalho, C; Bainbridge, M; Kannry, J

2013-01-01

Issues related to lack of system usability and potential safety hazards continue to be reported in the health information technology (HIT) literature. Usability engineering methods are increasingly used to ensure improved system usability and they are also beginning to be applied more widely for ensuring the safety of HIT applications. These methods are being used in the design and implementation of many HIT systems. In this paper we describe evidence-based approaches to applying usability engineering methods. A multi-phased approach to ensuring system usability and safety in healthcare is described. Usability inspection methods are first described including the development of evidence-based safety heuristics for HIT. Laboratory-based usability testing is then conducted under artificial conditions to test if a system has any base level usability problems that need to be corrected. Usability problems that are detected are corrected and then a new phase is initiated where the system is tested under more realistic conditions using clinical simulations. This phase may involve testing the system with simulated patients. Finally, an additional phase may be conducted, involving a naturalistic study of system use under real-world clinical conditions. The methods described have been employed in the analysis of the usability and safety of a wide range of HIT applications, including electronic health record systems, decision support systems and consumer health applications. It has been found that at least usability inspection and usability testing should be applied prior to the widespread release of HIT. However, wherever possible, additional layers of testing involving clinical simulations and a naturalistic evaluation will likely detect usability and safety issues that may not otherwise be detected prior to widespread system release. The framework presented in the paper can be applied in order to develop more usable and safer HIT, based on multiple layers of evidence.

Development of standard testing methods for nuclear-waste forms

International Nuclear Information System (INIS)

Mendel, J.E.; Nelson, R.D.

1981-11-01

Standard test methods for waste package component development and design, safety analyses, and licensing are being developed for the Nuclear Waste Materials Handbook. This paper describes mainly the testing methods for obtaining waste form materials data

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...

Validation testing of safety-critical software

International Nuclear Information System (INIS)

Kim, Hang Bae; Han, Jae Bok

1995-01-01

A software engineering process has been developed for the design of safety critical software for Wolsung 2/3/4 project to satisfy the requirements of the regulatory body. Among the process, this paper described the detail process of validation testing performed to ensure that the software with its hardware, developed by the design group, satisfies the requirements of the functional specification prepared by the independent functional group. To perform the tests, test facility and test software were developed and actual safety system computer was connected. Three kinds of test cases, i.e., functional test, performance test and self-check test, were programmed and run to verify each functional specifications. Test failures were feedback to the design group to revise the software and test results were analyzed and documented in the report to submit to the regulatory body. The test methodology and procedure were very efficient and satisfactory to perform the systematic and automatic test. The test results were also acceptable and successful to verify the software acts as specified in the program functional specification. This methodology can be applied to the validation of other safety-critical software. 2 figs., 2 tabs., 14 refs. (Author)

Safety Test Program Summary SNAP 19 Pioneer Heat Source Safety Program

Energy Technology Data Exchange (ETDEWEB)

None,

1971-07-01

Sixteen heat source assemblies have been tested in support of the SNAP 19 Pioneer Safety Test Program. Seven were subjected to simulated reentry heating in various plasma arc facilities followed by impact on earth or granite. Six assemblies were tested under abort accident conditions of overpressure, shrapnel impact, and solid and liquid propellant fires. Three capsules were hot impacted under Transit capsule impact conditions to verify comparability of test results between the two similar capsule designs, thus utilizing both Pioneer and Transit Safety Test results to support the Safety Analysis Report for Pioneer. The tests have shown the fuel is contained under all nominal accident environments with the exception of minor capsule cracks under severe impact and solid fire environments. No catastrophic capsule failures occurred in this test which would release large quantities of fuel. In no test was fuel visible to the eye following impact or fire. Breached capsules were defined as those which exhibit thoria contamination on its surface following a test, or one which exhibited visible cracks in the post test metallographic analyses.

Validation test case generation based on safety analysis ontology

International Nuclear Information System (INIS)

Fan, Chin-Feng; Wang, Wen-Shing

2012-01-01

Highlights: ► Current practice in validation test case generation for nuclear system is mainly ad hoc. ► This study designs a systematic approach to generate validation test cases from a Safety Analysis Report. ► It is based on a domain-specific ontology. ► Test coverage criteria have been defined and satisfied. ► A computerized toolset has been implemented to assist the proposed approach. - Abstract: Validation tests in the current nuclear industry practice are typically performed in an ad hoc fashion. This study presents a systematic and objective method of generating validation test cases from a Safety Analysis Report (SAR). A domain-specific ontology was designed and used to mark up a SAR; relevant information was then extracted from the marked-up document for use in automatically generating validation test cases that satisfy the proposed test coverage criteria; namely, single parameter coverage, use case coverage, abnormal condition coverage, and scenario coverage. The novelty of this technique is its systematic rather than ad hoc test case generation from a SAR to achieve high test coverage.

Review of domestic and international experience on optimization of tests planning for safety related systems at NPP

International Nuclear Information System (INIS)

Skalozubov, V.I.; Komarov, Yu.A.; Kolykanov, V.N.; Kochneva, V.Yu.; Gablaya, T.V.

2009-01-01

There are represented the basic requirements of normative and operating documents on test periodicity of safety related systems at NPPs, sets out the theoretical methods of test optimization of the technical systems, and analyses foreign engineering methods for changing test periodicity of the NPP systems. Based on this review analyses further tasks are formulated for improvement of the methodical base of optimization of tests planning for safety related systems

Safety testing for LHC access system

CERN Document Server

Valentini, F; Ninin, P; Scibile, S

2008-01-01

In the domain of Safety Real-Time Systems the problem of testing represents always a big effort in terms of time, costs and efficiency to guarantee an adequate coverage degree. Exhaustive tests may, in fact, not be practicable for large and distributed systems. This paper describes the testing process followed during the validation of the CERN's LHC Access System [1], responsible for monitoring and preventing physical risks for the personnel accessing the underground areas. In the paper we also present a novel strategy for the testing problem, intended to drastically reduce the time for the test patterns generation and execution. In particular, we propose a methodology for blackbox testing that relies on the application of Model Checking techniques. Model Checking is a formal method from computer science, commonly adopted to prove correctness of system’s models through an automatic system’s state space exploration against some property formulas.

9 CFR 113.38 - Guinea pig safety test.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Guinea pig safety test. 113.38 Section... Standard Procedures § 113.38 Guinea pig safety test. The guinea pig safety test provided in this section... be injected either intramuscularly or subcutaneously into each of two guinea pigs and the animals...

Standard test method for guided bend test for ductility of welds

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers a guided bend test for the determination of soundness and ductility of welds in ferrous and nonferrous products. Defects, not shown by X rays, may appear in the surface of a specimen when it is subjected to progressive localized overstressing. This guided bend test has been developed primarily for plates and is not intended to be substituted for other methods of bend testing. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. Note 1—For additional information see Terminology E 6, and American Welding Society Standard D 1.1. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

A Web-based Alternative Non-animal Method Database for Safety Cosmetic Evaluations

OpenAIRE

Kim, Seung Won; Kim, Bae-Hwan

2016-01-01

Animal testing was used traditionally in the cosmetics industry to confirm product safety, but has begun to be banned; alternative methods to replace animal experiments are either in development, or are being validated, worldwide. Research data related to test substances are critical for developing novel alternative tests. Moreover, safety information on cosmetic materials has neither been collected in a database nor shared among researchers. Therefore, it is imperative to build and share a d...

Cadmium safety rod thermal tests

International Nuclear Information System (INIS)

Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

1992-01-01

Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

Standard Test Method for Testing Nonmetallic Seal Materials by Immersion in a Simulated Geothermal Test Fluid

CERN Document Server

American Society for Testing and Materials. Philadelphia

1985-01-01

1.1 This test method covers a procedure for a laboratory test for performing an initial evaluation (screening) of nonmetallic seal materials by immersion in a simulated geothermal test fluid. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific precautionary statements, see Section 6 and 11.7.

Module Testing Techniques for Nuclear Safety Critical Software Using LDRA Testing Tool

International Nuclear Information System (INIS)

Moon, Kwon-Ki; Kim, Do-Yeon; Chang, Hoon-Seon; Chang, Young-Woo; Yun, Jae-Hee; Park, Jee-Duck; Kim, Jae-Hack

2006-01-01

The safety critical software in the I and C systems of nuclear power plants requires high functional integrity and reliability. To achieve those requirement goals, the safety critical software should be verified and tested according to related codes and standards through verification and validation (V and V) activities. The safety critical software testing is performed at various stages during the development of the software, and is generally classified as three major activities: module testing, system integration testing, and system validation testing. Module testing involves the evaluation of module level functions of hardware and software. System integration testing investigates the characteristics of a collection of modules and aims at establishing their correct interactions. System validation testing demonstrates that the complete system satisfies its functional requirements. In order to generate reliable software and reduce high maintenance cost, it is important that software testing is carried out at module level. Module testing for the nuclear safety critical software has rarely been performed by formal and proven testing tools because of its various constraints. LDRA testing tool is a widely used and proven tool set that provides powerful source code testing and analysis facilities for the V and V of general purpose software and safety critical software. Use of the tool set is indispensable where software is required to be reliable and as error-free as possible, and its use brings in substantial time and cost savings, and efficiency

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)

Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

International Nuclear Information System (INIS)

Wootan, D.W.; Omberg, R.P.

2015-01-01

Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

Helium leak testing methods in nuclear applications

International Nuclear Information System (INIS)

Ahmad, Anis

2004-01-01

Helium mass-spectrometer leak test is the most sensitive leak test method. It gives very reliable and sensitive test results. In last few years application of helium leak testing has gained more importance due to increased public awareness of safety and environment pollution caused by number of growing chemical and other such industries. Helium leak testing is carried out and specified in most of the critical area applications like nuclear, space, chemical and petrochemical industries

Standard Test Methods for Wet Insulation Integrity Testing of Photovoltaic Modules

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 These test methods provide procedures to determine the insulation resistance of a photovoltaic (PV) module, i.e. the electrical resistance between the module's internal electrical components and its exposed, electrically conductive, non-current carrying parts and surfaces. 1.2 The insulation integrity procedures are a combination of wet insulation resistance and wet dielectric voltage withstand test procedures. 1.3 These procedures are similar to and reference the insulation integrity test procedures described in Test Methods E 1462, with the difference being that the photovoltaic module under test is immersed in a wetting solution during the procedures. 1.4 These test methods do not establish pass or fail levels. The determination of acceptable or unacceptable results is beyond the scope of these test methods. 1.5 The values stated in SI units are to be regarded as the standard. 1.6 There is no similar or equivalent ISO standard. 1.7 This standard does not purport to address all of the safety conce...

Standard test method for calibration of surface/stress measuring devices

CERN Document Server

American Society for Testing and Materials. Philadelphia

1997-01-01

Return to Contents page 1.1 This test method covers calibration or verification of calibration, or both, of surface-stress measuring devices used to measure stress in annealed and heat-strengthened or tempered glass using polariscopic or refractometry based principles. 1.2 This test method is nondestructive. 1.3 This test method uses transmitted light, and therefore, is applicable to light-transmitting glasses. 1.4 This test method is not applicable to chemically tempered glass. 1.5 Using the procedure described, surface stresses can be measured only on the “tin” side of float glass. 1.6 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Application of risk-based methods to inservice testing of check valves

Energy Technology Data Exchange (ETDEWEB)

Closky, N.B.; Balkey, K.R.; McAllister, W.J. [and others

1996-12-01

Research efforts have been underway in the American Society of Mechanical Engineers (ASME) and industry to define appropriate methods for the application of risk-based technology in the development of inservice testing (IST) programs for pumps and valves in nuclear steam supply systems. This paper discusses a pilot application of these methods to the inservice testing of check valves in the emergency core cooling system of Georgia Power`s Vogtle nuclear power station. The results of the probabilistic safety assessment (PSA) are used to divide the check valves into risk-significant and less-risk-significant groups. This information is reviewed by a plant expert panel along with the consideration of appropriate deterministic insights to finally categorize the check valves into more safety-significant and less safety-significant component groups. All of the more safety-significant check valves are further evaluated in detail using a failure modes and causes analysis (FMCA) to assist in defining effective IST strategies. A template has been designed to evaluate how effective current and emerging tests for check valves are in detecting failures or in finding significant conditions that are precursors to failure for the likely failure causes. This information is then used to design and evaluate appropriate IST strategies that consider both the test method and frequency. A few of the less safety-significant check valves are also evaluated using this process since differences exist in check valve design, function, and operating conditions. Appropriate test strategies are selected for each check valve that has been evaluated based on safety and cost considerations. Test strategies are inferred from this information for the other check valves based on similar check valve conditions. Sensitivity studies are performed using the PSA model to arrive at an overall IST program that maintains or enhances safety at the lowest achievable cost.

Standard Test Method for Hydrophobic Surface Films by the Atomizer Test

CERN Document Server

American Society for Testing and Materials. Philadelphia

1965-01-01

1.1 This test method covers the detection of the presence of hydrophobic (nonwetting) films on surfaces and the presence of hydrophobic organic materials in processing ambients. When properly conducted, the test will enable detection of fractional molecular layers of hydrophobic organic contaminants. On very rough or porous surfaces the sensitivity of the test may be significantly decreased. 1.2 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Expert opinions on the acceptance of alternative methods in food safety evaluations

NARCIS (Netherlands)

Punt, Ans; Bouwmeester, Hans; Schiffelers, Marie Jeanne W.A.; Peijnenburg, Ad A.C.M.

2018-01-01

Inclusion of alternative methods that replace, reduce, or refine (3R) animal testing within regulatory safety evaluations of chemicals generally faces many hurdles. The goal of the current work is to i) collect responses from key stakeholders involved in food safety evaluations on what they consider

Thermal-hydraulic tests for reactor safety system

International Nuclear Information System (INIS)

Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

2002-05-01

Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

Standard test method for measuring pH of soil for use in corrosion testing

CERN Document Server

American Society for Testing and Materials. Philadelphia

1995-01-01

1.1 This test method covers a procedure for determining the pH of a soil in corrosion testing. The principle use of the test is to supplement soil resistivity measurements and thereby identify conditions under which the corrosion of metals in soil may be accentuated (see G 57 - 78 (1984)). 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Performance Testing Methodology for Safety-Critical Programmable Logic Controller

International Nuclear Information System (INIS)

Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

2009-01-01

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

Utilizing the Fast Flux Test Facility for international passive safety testing

International Nuclear Information System (INIS)

Shen, P.K.; Padilla, A.; Lucoff, D.M.; Waltar, A.E.

1991-01-01

A two-phased approach has been undertaken in the Fast Flux Test Facility (FFTF) to conduct passive safety testing. Phase I (1986 to 1987) was structured to obtain an initial understanding of the reactivity feedback components. The planned Phase II (1992 to 1993) international program will extend the testing to include static and dynamic feedback measurements, transient and demonstration tests, and gas expansion module (GEM) reactivity tests. The primary objective is to meet the needs for safety analysis code validation, with particular emphasis on reducing the uncertainties associated with structure reactivity feedback. Program scope and predicted FFTF responses are discussed and illustrated. (author)

Standard test methods for characterizing duplex grain sizes

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 These test methods provide simple guidelines for deciding whether a duplex grain size exists. The test methods separate duplex grain sizes into one of two distinct classes, then into specific types within those classes, and provide systems for grain size characterization of each type. 1.2 Units—The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety concerns associated with its use. It is the responsibility of the user of this standard to consult appropriate safety and health practices and determine the applicability of regulatory limitations prior to its use.

Safety of type and screen method compared to conventional antiglobulin crossmatch procedures for compatibility testing in Indian setting

Directory of Open Access Journals (Sweden)

Chaudhary Rajendra

2011-01-01

Full Text Available Background: Over the past 30 years, pretransfusion tests have undergone considerable modification. In 1984, AABB recommended that the full cross match could be replaced by an abbreviated cross match in patients with negative antibody screen. However, before implementation of such a policy, issue regarding safety of T & S needs to be evaluated. Objectives: The aim of pretransfusion testing (PTT is to ensure that enough red blood cells (RBCs in the selected red cell components will survive when transfused. Results and Conclusion: We have, therefore in this study; evaluated safety of T & S procedure for PTT in comparison with conventional test tube cross match. The T & S procedure gave a safety of 91.6%. Also, the usefulness of the T & S was shown through the detection of unexpected antibodies in 0.75% (15 out of 2026 of cases.

77 FR 4544 - CPSC Symposium on Phthalates Screening and Testing Methods

Science.gov (United States)

2012-01-30

... Screening and Testing Methods AGENCY: Consumer Product Safety Commission. ACTION: Notice. SUMMARY: The... symposium on phthalates screening and testing methods. The symposium will be held at the CPSC's National... submit comments, identified by Docket No. CPSC-2012-0008, by any of the following methods: Electronic...

46 CFR 61.30-20 - Automatic control and safety tests.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Automatic control and safety tests. 61.30-20 Section 61... TESTS AND INSPECTIONS Tests and Inspections of Fired Thermal Fluid Heaters § 61.30-20 Automatic control and safety tests. Operational tests and checks of all safety and limit controls, combustion controls...

Standard Test Method for Testing Polymeric Seal Materials for Geothermal and/or High Temperature Service Under Sealing Stress

CERN Document Server

American Society for Testing and Materials. Philadelphia

1985-01-01

1.1 This test method covers the initial evaluation of (screening) polymeric materials for seals under static sealing stress and at elevated temperatures. 1.2 This test method applies to geothermal service only if used in conjunction with Test Method E 1068. 1.3 The test fluid is distilled water. 1.4 The values stated in SI units are to be regarded as the standard. The values in parentheses are for information only. 1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Fire Safety Tests for Spherical Resorcinol Formaldehyde Resin: Data Summary Report

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Sang; Peterson, Reid A.; Schweiger, Michael J.

2012-07-30

A draft safety evaluation of the scenario for spherical resorcinol-formaldehyde (SRF) resin fire inside the ion exchange column was performed by the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Fire Safety organization. The result of this draft evaluation suggested a potential change of the fire safety classification for the Cesium Ion Exchange Process System (CXP) emergency elution vessels, equipment, and piping, which may be overly bounding based on the fire performance data from the manufacturer of the ion exchange resin selected for use at the WTP. To resolve this question, the fire properties of the SRF resin were measured by Southwest Research Institute (SwRI), following the American Society for Testing and Materials (ASTM) standard procedures, through a subcontract managed by Pacific Northwest National Laboratory (PNNL). For some tests, the ASTM standard procedures were not entirely appropriate or practical for the SRF resin material, so the procedures were modified and deviations from the ASTM standard procedures were noted. This report summarizes the results of fire safety tests performed and reported by SwRI. The efforts by PNNL were limited to summarizing the test results provided by SwRI into one consolidated data report. All as-received SwRI reports are attached to this report in the Appendix. Where applicable, the precision and bias of each test method, as given by each ASTM standard procedure, are included and compared with the SwRI test results of the SRF resin.

Model-based testing for software safety

NARCIS (Netherlands)

Gurbuz, Havva Gulay; Tekinerdogan, Bedir

2017-01-01

Testing safety-critical systems is crucial since a failure or malfunction may result in death or serious injuries to people, equipment, or environment. An important challenge in testing is the derivation of test cases that can identify the potential faults. Model-based testing adopts models of a

Standard Test Method for Solar Photometric Transmittance of Sheet Materials Using Sunlight

CERN Document Server

American Society for Testing and Materials. Philadelphia

1996-01-01

1.1 This test method covers the measurement of solar photometric transmittance of materials in sheet form. Solar photometric transmittance is measured using a photometer (illuminance meter) in an enclosure with the sun and sky as the source of radiation. The enclosure and method of test is specified in Test Method E 1175 (or Test Method E 1084). 1.2 The purpose of this test method is to specify a photometric sensor to be used with the procedure for measuring the solar photometric transmittance of sheet materials containing inhomogeneities in their optical properties. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Comparison of EPRI safety valve test data with analytically determined hydraulic results

International Nuclear Information System (INIS)

Smith, L.C.; Howe, K.S.

1983-01-01

NUREG-0737 (November 1980) and all subsequent U.S. NRC generic follow-up letters require that all operating plant licensees and applicants verify the acceptability of plant specific pressurizer safety valve piping systems for valve operation transients by testing. To aid in this verification process, the Electric Power Research Institute (EPRI) conducted an extensive testing program at the Combustion Engineering Test Facility. Pertinent tests simulating dynamic opening of the safety valves for representative upstream environments were carried out. Different models and sizes of safety valves were tested at the simulated operating conditions. Transducers placed at key points in the system monitored a variety of thermal, hydraulic and structural parameters. From this data, a more complete description of the transient can be made. The EPRI test configuration was analytically modeled using a one-dimensional thermal hydraulic computer program that uses the method of characteristics approach to generate key fluid parameters as a function of space and time. The conservative equations are solved by applying both the implicit and explicit characteristic methods. Unbalanced or wave forces were determined for each straight run of pipe bounded on each side by a turn or elbow. Blowdown forces were included, where appropriate. Several parameters were varied to determine the effects on the pressure, hydraulic forces and timings of events. By comparing these quantities with the experimentally obtained data, an approximate picture of the flow dynamics is arrived at. Two cases in particular are presented. These are the hot and cold loop seal discharge tests made with the Crosby 6M6 spring-loaded safety valve. Included in the paper is a description of the hydraulic code, modeling techniques and assumptions, a comparison of the numerical results with experimental data and a qualitative description of the factors which govern pipe support loading. (orig.)

Standard Test Method for Hydrophobic Surface Films by the Water-Break Test

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers the detection of the presence of hydrophobic (nonwetting) films on surfaces and the presence of hydrophobic organic materials in processing ambients. When properly conducted, the test will enable detection of molecular layers of hydrophobic organic contaminants. On very rough or porous surfaces, the sensitivity of the test may be significantly decreased. 1.2 The values stated in SI units are to be regarded as the standard. The inch-pound values given in parentheses are for information only. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Safety relief valve alternate analysis method

International Nuclear Information System (INIS)

Adams, R.H.; Javid, A.; Khatua, T.P.

1981-01-01

An experimental test program was started in the United States in 1976 to define and quantify Safety Relief Valve (SRV) phenomena in General Electric Mark I Suppression Chambers. The testing considered several discharged devices and was used to correlate SRV load prediction models. The program was funded by utilities with Mark I containments and has resulted in a detailed SRV load definition as a portion of the Mark I containment program Load Definition Report (LDR). The (USNRC) has reviewed and approved the LDR SRV load definition. In addition, the USNRC has permitted calibration of structural models used for predicting torus response to SRV loads. Model calibration is subject to confirmatory in-plant testing. The SRV methodology given in the LDR requires that transient dynamic pressures be applied to a torus structural model that includes a fluid added mass matrix. Preliminary evaluations of torus response have indicated order of magnitude conservatisms, with respect to test results, which could result in unrealistic containment modifications. In addition, structural response trends observed in full-scale tests between cold pipe, first valve actuation and hot pipe, subsequent valve actuation conditions have not been duplicated using current analysis methods. It was suggested by others that an energy approach using current fluid models be utilized to define loads. An alternate SRV analysis method is defined to correct suppression chamber structural response to a level that permits economical but conservative design. Simple analogs are developed for the purpose of correcting the analytical response obtained from LDR analysis methods. Analogs evaluated considered forced vibration and free vibration structural response. The corrected response correlated well with in-plant test response. The correlation of the analytical model at test conditions permits application of the alternate analysis method at design conditions. (orig./HP)

Standard test method for exfoliation corrosion susceptibility in 2XXX and 7XXX Series Aluminum Alloys (EXCO Test)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 This test method covers a procedure for constant immersion exfoliation corrosion (EXCO) testing of high-strength 2XXX and 7XXX series aluminum alloys. Note 1—This test method was originally developed for research and development purposes; however, it is referenced, in specific material specifications, as applicable for evaluating production material (refer to Section 14 on Precision and Bias). 1.2 This test method applies to all wrought products such as sheet, plate, extrusions, and forgings produced from conventional ingot metallurgy process. 1.3 This test method can be used with any form of specimen or part that can be immersed in the test solution. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Review of design criteria and safety analysis of safety class electric building for fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1998-02-01

Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

[Non-animal toxicology in the safety testing of chemicals].

Science.gov (United States)

Heinonen, Tuula; Tähti, Hanna

2013-01-01

There is an urgent need to develop predictive test methods better than animal experiments for assessing the safety of chemical substances to man. According to today's vision this is achieved by using human cell based tissue and organ models. In the new testing strategy the toxic effects are assessed by the changes in the critical parameters of the cellular biochemical routes (AOP, adverse toxic outcome pathway-principle) in the target tissues. In vitro-tests are rapid and effective, and with them automation can be applied. The change in the testing paradigm is supported by all stakeholders: scientists, regulators and people concerned on animal welfare.

French safety and criticality testing programmes

International Nuclear Information System (INIS)

Barbry, F.; Leclerc, J.; Manaranche, J.C.; Maubert, L.

1982-01-01

This article underlines the need to include experimental safety-criticality programmes in the French nuclear effort. The means and methods used at the Section of Experimental Nuclear Safety and Criticality Research, attached to the CEA Valduc Centre, are described. Three experimental programmes are presented: safety-criticality of the PWR fuel cycle, neutron poisoning of plutonium solutions by gadolinium and safety-criticality of slightly enriched and slightly moderated uranium oxide. Criticality accidents studies in solution are then described [fr

Optimisation of test and maintenance based on probabilistic methods

International Nuclear Information System (INIS)

Cepin, M.

2001-01-01

This paper presents a method, which based on models and results of probabilistic safety assessment, minimises the nuclear power plant risk by optimisation of arrangement of safety equipment outages. The test and maintenance activities of the safety equipment are timely arranged, so the classical static fault tree models are extended with the time requirements to be capable to model real plant states. A house event matrix is used, which enables modelling of the equipment arrangements through the discrete points of time. The result of the method is determination of such configuration of equipment outages, which result in the minimal risk. Minimal risk is represented by system unavailability. (authors)

Method for selection of optimal road safety composite index with examples from DEA and TOPSIS method.

Science.gov (United States)

Rosić, Miroslav; Pešić, Dalibor; Kukić, Dragoslav; Antić, Boris; Božović, Milan

2017-01-01

Concept of composite road safety index is a popular and relatively new concept among road safety experts around the world. As there is a constant need for comparison among different units (countries, municipalities, roads, etc.) there is need to choose an adequate method which will make comparison fair to all compared units. Usually comparisons using one specific indicator (parameter which describes safety or unsafety) can end up with totally different ranking of compared units which is quite complicated for decision maker to determine "real best performers". Need for composite road safety index is becoming dominant since road safety presents a complex system where more and more indicators are constantly being developed to describe it. Among wide variety of models and developed composite indexes, a decision maker can come to even bigger dilemma than choosing one adequate risk measure. As DEA and TOPSIS are well-known mathematical models and have recently been increasingly used for risk evaluation in road safety, we used efficiencies (composite indexes) obtained by different models, based on DEA and TOPSIS, to present PROMETHEE-RS model for selection of optimal method for composite index. Method for selection of optimal composite index is based on three parameters (average correlation, average rank variation and average cluster variation) inserted into a PROMETHEE MCDM method in order to choose the optimal one. The model is tested by comparing 27 police departments in Serbia. Copyright © 2016 Elsevier Ltd. All rights reserved.

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)

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.

A New Method for the Evaluation of Vaccine Safety Based on Comprehensive Gene Expression Analysis

Directory of Open Access Journals (Sweden)

Haruka Momose

2010-01-01

Full Text Available For the past 50 years, quality control and safety tests have been used to evaluate vaccine safety. However, conventional animal safety tests need to be improved in several aspects. For example, the number of test animals used needs to be reduced and the test period shortened. It is, therefore, necessary to develop a new vaccine evaluation system. In this review, we show that gene expression patterns are well correlated to biological responses in vaccinated rats. Our findings and methods using experimental biology and genome science provide an important means of assessment for vaccine toxicity.

Standard Test Method for Determining the Linearity of a Photovoltaic Device Parameter with Respect To a Test Parameter

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 This test method determines the degree of linearity of a photovoltaic device parameter with respect to a test parameter, for example, short-circuit current with respect to irradiance. 1.2 The linearity determined by this test method applies only at the time of testing, and implies no past or future performance level. 1.3 This test method applies only to non-concentrator terrestrial photovoltaic devices. 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Method for Solar Transmittance (Terrestrial) of Sheet Materials Using Sunlight

CERN Document Server

American Society for Testing and Materials. Philadelphia

1986-01-01

1.1 This test method covers the measurement of solar transmittance (terrestrial) of materials in sheet form by using a pyranometer, an enclosure, and the sun as the energy source. 1.2 This test method also allows measurement of solar transmittance at angles other than normal incidence. 1.3 This test method is applicable to sheet materials that are transparent, translucent, textured, or patterned. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Selenide isotope generator for the Galileo Mission: safety test plan

International Nuclear Information System (INIS)

1979-01-01

The intent of this safety test plan is to outline particular kinds of safety tests designed to produce information which would be useful in the safety analysis process. The program deals primarily with the response of the RTG to accident environments; accordingly two criteria were established: (1) safety tests should be performed for environments which are the most critical in terms of risk contribution; and (2) tests should be formulated to determine failure conditions for critical heat source components rather than observe heat source response in reference accident environments. To satisfy criterion 1. results of a recent safety study were used to rank various accidents in terms of expected source terms. Six kinds of tests were then proposed which would provide information meeting the second criterion

Standard test method for measurement of soil resistivity using the two-electrode soil box method

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 This test method covers the equipment and a procedure for the measurement of soil resistivity, for samples removed from the ground, for use in the control of corrosion of buried structures. 1.2 Procedures allow for this test method to be used n the field or in the laboratory. 1.3 The test method procedures are for the resistivity measurement of soil samples in the saturated condition and in the as-received condition. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. Soil resistivity values are reported in ohm-centimeter. 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 appropriate safety and health practices and to determine the applicability of regulatory limitations prior to use.

PWR core safety analysis with 3-dimensional methods

International Nuclear Information System (INIS)

Gensler, A.; Kühnel, K.; Kuch, S.

2015-01-01

Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

Standard test method for measuring waste glass or glass ceramic durability by vapor hydration test

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 The vapor hydration test method can be used to study the corrosion of a waste forms such as glasses and glass ceramics upon exposure to water vapor at elevated temperatures. In addition, the alteration phases that form can be used as indicators of those phases that may form under repository conditions. These tests; which allow altering of glass at high surface area to solution volume ratio; provide useful information regarding the alteration phases that are formed, the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (McGrail et al, 2002 (1) for example). 1.2 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data. 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 appropriate safety and health practice...

Standard test methods for elevated temperature tension tests of metallic materials

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 These test methods cover procedure and equipment for the determination of tensile strength, yield strength, elongation, and reduction of area of metallic materials at elevated temperatures. 1.2 Determination of modulus of elasticity and proportional limit are not included. 1.3 Tension tests under conditions of rapid heating or rapid strain rates are not included. 1.4 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Passive safety testing at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Lucoff, D.M.

1989-01-01

During 1986, the Fast Flux Test Facility (FFTF) conducted several tests designed to improve the understanding of the passive safety characteristics of an oxide-fueled liquid-metal reactor (LMR). Static and dynamic tests were performed over a broad range of power, flow, and temperature conditions that extended beyond those for normal operation. Key results of these tests are presented. Stable operation at low power with natural circulation cooling was demonstrated. A passive safety enhancement feature, the gas expansion module (GEM) was developed specifically to offset the large amount of cooldown reactivity that needs to be controlled in an oxide-fueled LMR undergoing an unprotected loss-of-flow accident. Nine GEMs were built and successfully tested in FFTF. With the reactor at 50% power (200 MW (thermal)), the main coolant pumps were turned off and the normal control rod scram response was inhibited. The GEMs and inherent core reactivity feedback mechanisms took the core subcritical with a modest peak coolant temperature transient that reached 85 degrees C above the pretransient value and always maintained a >400 degrees C margin to the sodium boiling point (910 degrees C)

Standard test method for conducting drop-weight test to determine nil-ductility transition temperature of ferritic steels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2006-01-01

1.1 This test method covers the determination of the nil-ductility transition (NDT) temperature of ferritic steels, 5/8 in. (15.9 mm) and thicker. 1.2 This test method may be used whenever the inquiry, contract, order, or specification states that the steels are subject to fracture toughness requirements as determined by the drop-weight test. 1.3 The values stated in inch-pound units are to be regarded as the standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

16 CFR 1509.6 - Component-spacing test method.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Component-spacing test method. 1509.6 Section 1509.6 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION FEDERAL HAZARDOUS SUBSTANCES ACT... applied to the wedge perpendicular to the plane of the crib side. ...

Standard test method for wear testing with a pin-on-disk apparatus

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 This test method covers a laboratory procedure for determining the wear of materials during sliding using a pin-on-disk apparatus. Materials are tested in pairs under nominally non-abrasive conditions. The principal areas of experimental attention in using this type of apparatus to measure wear are described. The coefficient of friction may also be determined. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Extended occupant safety through virtual testing: objectives of the European project VITES

NARCIS (Netherlands)

Hoof, J.F.A.M. van; Happee, R.; Baldauf, H.; Puppini, R.; Oakley, C.

2001-01-01

Increasingly stringent international passenger safety norms and the need to reduce vehicle body weight for environmental and protection requirements demand efficient and innovative design methods. Computer simulation, or virtual testing, allows an integrated evaluation of these aspects in the early

Relevance of microbial finished product testing in food safety management

DEFF Research Database (Denmark)

Zwietering, Marcel H.; Jacxsens, Liesbeth; Membré, Jeanne Marie

2016-01-01

Management of microbiological food safety is largely based on good design of processes, products and procedures. Finished product testing may be considered as a control measure at the end of the production process. However, testing gives only very limited information on the safety status of a food......-active way by implementing an effective food safety management system. For verification activities in a food safety management system, finished product testing may however be useful. For three cases studies; canned food, chocolate and cooked ham, the relevance of testing both of finished products....... If a hazardous organism is found it means something, but absence in a limited number of samples is no guarantee of safety of a whole production batch. Finished product testing is often too little and too late. Therefore most attention should be focussed on management and control of the hazards in a more pro...

Routine testing on protective and safety systems and components

International Nuclear Information System (INIS)

Rysy, W.

1977-01-01

1) In-process inspection, tests during commissioning. 2) Tests during reactor operation. 2.1) Reactor protection system, for example: continuous auto-testing by a dynamic system, check of the output signals; 2.2) safety features: selected examples: functional tests on the ECCS, trial operation of the emergency diesels. 3) Tests during refuelling phase. 3.1) Containment: Leakage rate tests, leak testing; 3.2) coolant system: selected examples: inservice inspections of the pressure vessel, eddy current testing of the steam generator, functional tests of safety valves. (orig./HP) [de

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).

Method of accounting for code safety valve setpoint drift in safety analyses

International Nuclear Information System (INIS)

Rousseau, K.R.; Bergeron, P.A.

1989-01-01

In performing the safety analyses for transients that result in a challenge to the reactor coolant system (RCS) pressure boundary, the general acceptance criterion is that the peak RCS pressure not exceed the American Society of Mechanical Engineers limit of 110% of the design pressure. Without crediting non-safety-grade pressure mitigating systems, protection from this limit is mainly provided by the primary and secondary code safety valves. In theory, the combination of relief capacity and setpoints for these valves is designed to provide this protection. Generally, banks of valves are set at varying setpoints staggered by 15- to 20-psid increments to minimize the number of valves that would open by an overpressure challenge. In practice, however, when these valves are removed and tested (typically during a refueling outage), setpoints are sometimes found to have drifted by >50 psid. This drift should be accounted for during the performance of the safety analysis. This paper describes analyses performed by Yankee Atomic Electric Company (YAEC) to account for setpoint drift in safety valves from testing. The results of these analyses are used to define safety valve operability or acceptance criteria

77 FR 2606 - Pipeline Safety: Random Drug Testing Rate

Science.gov (United States)

2012-01-18

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID PHMSA-2012-0004] Pipeline Safety: Random Drug Testing Rate AGENCY: Pipeline and Hazardous Materials... pipelines and operators of liquefied natural gas facilities must select and test a percentage of covered...

75 FR 9018 - Pipeline Safety: Random Drug Testing Rate

Science.gov (United States)

2010-02-26

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID PHMSA-2010-0034] Pipeline Safety: Random Drug Testing Rate AGENCY: Pipeline and Hazardous Materials... pipelines and operators of liquefied natural gas facilities must select and test a percentage of covered...

Safety report content and development for test loop facility on MARIA reactor

International Nuclear Information System (INIS)

Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

1982-01-01

A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

Revision of nuclear power plants safety systems' routine testing assigned periodicity during the design extension period

International Nuclear Information System (INIS)

Skalozubov, V.I.; Kozlov, Yi.L.; Chulkyin, O.O.

2017-01-01

When nuclear power plants safety systems thermal equipment operation extending, a necessary requirement shall rely on revising the scheduled equipment tests frequency to optimize those tests schedule taking into account the equipment remained lifespan. On the one hand, there exists a need for tests frequency increase to detect ''hidden'' failures, and on the another, frequent tests cause a premature wear of the equipment. Proposed is an original method for optimizing the frequency of NPPs safety systems thermal engineering equipment testing. Essential in the proposed method is the optimization criterion chosen: index of security system failure probability non-exceedance during the beyond-design operating period as referred to the failure probability expected considering the equipment residual resource during the design operating period. The developed method implementation when applied to NPPs safety systems operated beyond the design service life at nuclear power plants with WWER-1000 series reactors, allowed to establish that the optimal tests frequency makes half the designed one when the equipment service life is extended by five years and three times less that the designed frequency when subject lifespan extended by 10 years.

Verification of FPGA-Signal using the test board which is applied to Safety-related controller

Energy Technology Data Exchange (ETDEWEB)

Chung, Youn-Hu; Yoo, Kwanwoo; Lee, Myeongkyun; Yun, Donghwa [SOOSAN ENS, Seoul (Korea, Republic of)

2016-10-15

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. Saftety-related PLC must meet the international standard specifications. With this reason, we use V and V according to an international standard in order to secure high reliability and safety. By using this, we are supposed to proceed to a variety of verification courses for extra reliability and safety analysis. In order to have efficient verification of test results, we propose the test using the newly changed BGA socket which can resolve the problems of the conventional socket on this paper. The Verification of processes is divided into verification of Hardware and firmware. That processes are carried out in the unit testing and integration testing. The proposed test method is simple, the effect of cost reductions by batch process. In addition, it is advantageous to measure the signal from the Hi-speed-IC due to its short length of the pins and it was plated with the copper around it. Further, it also to prevent abrasion on the IC ball because it has no direct contact with the PCB. Therefore, it can be actually applied is to the BGA package test and we can easily verify logic as well as easily checking the operation of the designed data.

Study on personnel qualification for non-destructive tests in the field of reactor safety

International Nuclear Information System (INIS)

Trusch, K.; Wuestenberg, H.

1977-01-01

The training system for non-destructive testing is described, and the available and necessary personnel is analyzed; the personnel required for reactor safety problems is treated separately. On this basis, the subjects discussed in the study - available personnel, personnel requirements, training, training requirements, and suggestions for realisation - are treated in a general manner to begin with and afterwards with a view to specific problems of reactor safety. The methods employed are adapted to this situation. To obtain the necessary empirical data, questionnaires were set up and distributed, and experts in selected business companies and institutions were interviewed who work in the field of reactor safety or do same training in non-destructive testing. (orig.) [de

Feasibility studies of safety assessment methods for programmable automation systems. Final report of the AVV project

International Nuclear Information System (INIS)

Haapanen, P.; Maskuniitty, M.; Pulkkinen, U.; Heikkinen, J.; Korhonen, J.; Tuulari, E.

1995-10-01

Feasibility studies of two different groups of methodologies for safety assessment of programmable automation systems has been executed at the Technical Research Centre of Finland (VTT). The studies concerned the dynamic testing methods and the fault tree (FT) and failure mode and effects analysis (FMEA) methods. In order to get real experience in the application of these methods, an experimental testing of two realistic pilot systems were executed and a FT/FMEA analysis of a programmable safety function accomplished. The purpose of the studies was not to assess the object systems, but to get experience in the application of methods and assess their potentials and development needs. (46 refs., 21 figs.)

Field Test of the World Health Organization Multi-Professional Patient Safety Curriculum Guide

Science.gov (United States)

Farley, Donna; Zheng, Hao; Rousi, Eirini; Leotsakos, Agnès

2015-01-01

Introduction Although the importance of training in patient safety has been acknowledged for over a decade, it remains under-utilized and under-valued in most countries. WHO developed the Multi-professional Patient Safety Curriculum Guide to provide schools with the requirements and tools for incorporating patient safety in education. It was field tested with 12 participating schools across the six WHO regions, to assess its effectiveness for teaching patient safety to undergraduate and graduate students in a global variety of settings. Methods The evaluation used a combined prospective/retrospective design to generate formative information on the experiences of working with the Guide and summative information on the impacts of the Guide. Using stakeholder interviews and student surveys, data were gathered from each participating school at three times: the start of the field test (baseline), soon after each school started teaching, and soon after each school finished teaching. Results Stakeholders interviewed were strongly positive about the Guide, noting that it emphasized universally important patient safety topics, was culturally appropriate for their countries, and gave credibility and created a focus on patient safety at their schools. Student perceptions and attitudes regarding patient safety improved substantially during the field test, and their knowledge of the topics they were taught doubled, from 10.7% to 20.8% of correct answers on the student survey. Discussion This evaluation documented the effectiveness of the Curriculum Guide, for both ease of use by schools and its impacts on improving the patient safety knowledge of healthcare students. WHO should be well positioned to refine the contents of the Guide and move forward in encouraging broader use of the Guide globally for teaching patient safety. PMID:26406893

Large-Scale Spacecraft Fire Safety Tests

Science.gov (United States)

Urban, David; Ruff, Gary A.; Ferkul, Paul V.; Olson, Sandra; Fernandez-Pello, A. Carlos; T'ien, James S.; Torero, Jose L.; Cowlard, Adam J.; Rouvreau, Sebastien; Minster, Olivier;

Standard Test Method for Gel Time of Carbon Fiber-Epoxy Prepreg

CERN Document Server

American Society for Testing and Materials. Philadelphia

1999-01-01

1.1 This test method covers the determination of gel time of carbon fiber-epoxy tape and sheet. The test method is suitable for the measurement of gel time of resin systems having either high or low viscosity. 1.2 The values stated in SI units are to be regarded as standard. The values in parentheses are for reference only. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard test method for determining atmospheric chloride deposition rate by wet candle method

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers a wet candle device and its use in measuring atmospheric chloride deposition (amount of chloride salts deposited from the atmosphere on a given area per unit time). 1.2 Data on atmospheric chloride deposition can be useful in classifying the corrosivity of a specific area, such as an atmospheric test site. Caution must be exercised, however, to take into consideration the season because airborne chlorides vary widely between seasons. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Use of the bootstrap method to develop a physical fitness test for public safety officers who serve as both police officers and firefighters.

Science.gov (United States)

Adams, Jenny; Cheng, Dunlei; Lee, John; Shock, Tiffany; Kennedy, Kathleen; Pate, Scotty

2014-07-01

Physical fitness testing is a common tool for motivating employees with strenuous occupations to reach and maintain a minimum level of fitness. Nevertheless, the use of such tests can be hampered by several factors, including required compliance with US antidiscrimination laws. The Highland Park (Texas) Department of Public Safety implemented testing in 1991, but no single test adequately evaluated its sworn employees, who are cross-trained and serve as police officers and firefighters. In 2010, the department's fitness experts worked with exercise physiologists from Baylor Heart and Vascular Hospital to develop and evaluate a single test that would be equitable regardless of race/ethnicity, disability, sex, or age >50 years. The new test comprised a series of exercises to assess overall fitness, followed by two sequences of job-specific tasks related to firefighting and police work, respectively. The study group of 50 public safety officers took the test; raw data (e.g., the number of repetitions performed or the time required to complete a task) were collected during three quarterly testing sessions. The statistical bootstrap method was then used to determine the levels of performance that would correlate with 0, 1, 2, or 3 points for each task. A sensitivity analysis was done to determine the overall minimum passing score of 17 points. The new physical fitness test and scoring system have been incorporated into the department's policies and procedures as part of the town's overall employee fitness program.

Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

International Nuclear Information System (INIS)

Wright, A.E.; Dutt, D.S.; Harrison, L.J.

1990-01-01

Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

Evaluation of safety test needs for the gas cooled breeder reactors

International Nuclear Information System (INIS)

Emon, D.E.; Buttemer, D.R.; Sevy, R.H.

1976-01-01

This paper deals with the process used in determining the safety test needs for the Gas Cooled Fast Breeder Reactor (GCFR), reports existing tentative conclusions, and indicates the direction that the process is taking at this time. The process is based upon two ideas: (1) that the safety information needs will be identified through risk analysis directly dependent on the various design features of the GCFR and (2) that the safety program will be determined by a safety review committee. The paper limits itself to presenting thoughts on the safety test needs directly associated with the GCFR core during severe beyond design basis accident situations involving the loss of coolable core geometry. Representative event sequence diagrams are reported for the three generic classes of accidents considered. The following categories of information are identified: safety information needs, safety tests required to fulfill these information needs, and the facilities required to perform the tests

Study of test methods for radionuclide migration in aerated zone

International Nuclear Information System (INIS)

Li Shushen; Guo Zede; Wang Zhiming

1993-01-01

Aerated zone is an important natural barrier against transport of radionuclides released from disposal facilities of LLRW. This paper introduces study methods for radionuclide migration in aerated zone, including determination of water movement, laboratory simulation test, and field tracing test. For one purpose, results obtained with different methods are compared. These methods have been used in a five-year cooperative research project between CIRP and JAERI for an establishment of methodology for safety assessment on shallow land disposal of LLRW

Methods and strategies for future reactor safety goals

Science.gov (United States)

Arndt, Steven Andrew

-informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than

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

Standard test method for radiochemical determination of plutonium in Soil by alpha spectroscopy

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 This test method covers the determination of plutonium in soils at levels of detection dependent on count time, sample size, detector, background, and tracer yield. This test method describes one acceptable approach to the determination of plutonium in soil. 1.2 This test method is designed for 10 g of soil, previously collected and treated as described in Practices C998 and C999, but sample sizes up to 50 g may be analyzed by this test method. This test method may not be able to completely dissolve all forms of plutonium in the soil matrix. 1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are for information only. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Specific hazard statements are given in Section 9.

Cross-comparison of three surrogate safety methods to diagnose cyclist safety problems at intersections in Norway.

Science.gov (United States)

Laureshyn, Aliaksei; Goede, Maartje de; Saunier, Nicolas; Fyhri, Aslak

2017-08-01

Relying on accident records as the main data source for studying cyclists' safety has many drawbacks, such as high degree of under-reporting, the lack of accident details and particularly of information about the interaction processes that led to the accident. It is also an ethical problem as one has to wait for accidents to happen in order to make a statement about cyclists' (un-)safety. In this perspective, the use of surrogate safety measures based on actual observations in traffic is very promising. In this study we used video data from three intersections in Norway that were all independently analysed using three methods: the Swedish traffic conflict technique (Swedish TCT), the Dutch conflict technique (DOCTOR) and the probabilistic surrogate measures of safety (PSMS) technique developed in Canada. The first two methods are based on manual detection and counting of critical events in traffic (traffic conflicts), while the third considers probabilities of multiple trajectories for each interaction and delivers a density map of potential collision points per site. Due to extensive use of microscopic data, PSMS technique relies heavily on automated tracking of the road users in video. Across the three sites, the methods show similarities or are at least "compatible" with the accident records. The two conflict techniques agree quite well for the number, type and location of conflicts, but some differences with no obvious explanation are also found. PSMS reports many more safety-relevant interactions including less severe events. The location of the potential collision points is compatible with what the conflict techniques suggest, but the possibly significant share of false alarms due to inaccurate trajectories extracted from video complicates the comparison. The tested techniques still require enhancement, with respect to better adjustment to analysis of the situations involving cyclists (and vulnerable road users in general) and further validation. However, we

Posttest analysis of the FFTF inherent safety tests

International Nuclear Information System (INIS)

Padilla, A. Jr.; Claybrook, S.W.

1987-01-01

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

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.

Prediction of safety critical software operational reliability from test reliability using testing environment factors

International Nuclear Information System (INIS)

Jung, Hoan Sung; Seong, Poong Hyun

1999-01-01

It has been a critical issue to predict the safety critical software reliability in nuclear engineering area. For many years, many researches have focused on the quantification of software reliability and there have been many models developed to quantify software reliability. Most software reliability models estimate the reliability with the failure data collected during the test assuming that the test environments well represent the operation profile. User's interest is however on the operational reliability rather than on the test reliability. The experiences show that the operational reliability is higher than the test reliability. With the assumption that the difference in reliability results from the change of environment, from testing to operation, testing environment factors comprising the aging factor and the coverage factor are developed in this paper and used to predict the ultimate operational reliability with the failure data in testing phase. It is by incorporating test environments applied beyond the operational profile into testing environment factors. The application results show that the proposed method can estimate the operational reliability accurately. (Author). 14 refs., 1 tab., 1 fig

Human-system safety methods for development of advanced air traffic management systems

International Nuclear Information System (INIS)

Nelson, William R.

1999-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) is supporting the National Aeronautics and Space Administration in the development of advanced air traffic management (ATM) systems as part of the Advanced Air Transportation Technologies program. As part of this program INEEL conducted a survey of human-system safety methods that have been applied to complex technical systems, to identify lessons learned from these applications and provide recommendations for the development of advanced ATM systems. The domains that were surveyed included offshore oil and gas, commercial nuclear power, commercial aviation, and military. The survey showed that widely different approaches are used in these industries, and that the methods used range from very high-level, qualitative approaches to very detailed quantitative methods such as human reliability analysis (HRA) and probabilistic safety assessment (PSA). In addition, the industries varied widely in how effectively they incorporate human-system safety assessment in the design, development, and testing of complex technical systems. In spite of the lack of uniformity in the approaches and methods used, it was found that methods are available that can be combined and adapted to support the development of advanced air traffic management systems (author) (ml)

Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1989-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

Formal testing and utilization of streaming media to improve flight crew safety knowledge.

Science.gov (United States)

Bellazzini, Marc A; Rankin, Peter M; Quisling, Jason; Gangnon, Ronald; Kohrs, Mike

2008-01-01

Increased concerns over the safety of air medical transport have prompted development of novel ways to increase safety. The objective of our study was to determine if an Internet streaming media safety video increased crew safety knowledge. 23 out of 40 crew members took an online safety pre-test, watched a safety video specific to our program and completed immediate and long-term post-testing 6 months later. Mean pre-test, post-test and 6 month follow up test scores were 84.9%, 92.3% and 88.4% respectively. There was a statistically significant difference in all scores (p Streaming media proved to be an accessible and effective supplement to safety training in our study.

Standard Test Method for Preparing Aircraft Cleaning Compounds, Liquid Type, Water Base, for Storage Stability Testing

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers the determination of the stability in storage, of liquid, water-base chemical cleaning compounds, used to clean the exterior surfaces of aircraft. 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

Reliability Verification of DBE Environment Simulation Test Facility by using Statistics Method

International Nuclear Information System (INIS)

Jang, Kyung Nam; Kim, Jong Soeg; Jeong, Sun Chul; Kyung Heum

2011-01-01

In the nuclear power plant, all the safety-related equipment including cables under the harsh environment should perform the equipment qualification (EQ) according to the IEEE std 323. There are three types of qualification methods including type testing, operating experience and analysis. In order to environmentally qualify the safety-related equipment using type testing method, not analysis or operation experience method, the representative sample of equipment, including interfaces, should be subjected to a series of tests. Among these tests, Design Basis Events (DBE) environment simulating test is the most important test. DBE simulation test is performed in DBE simulation test chamber according to the postulated DBE conditions including specified high-energy line break (HELB), loss of coolant accident (LOCA), main steam line break (MSLB) and etc, after thermal and radiation aging. Because most DBE conditions have 100% humidity condition, in order to trace temperature and pressure of DBE condition, high temperature steam should be used. During DBE simulation test, if high temperature steam under high pressure inject to the DBE test chamber, the temperature and pressure in test chamber rapidly increase over the target temperature. Therefore, the temperature and pressure in test chamber continue fluctuating during the DBE simulation test to meet target temperature and pressure. We should ensure fairness and accuracy of test result by confirming the performance of DBE environment simulation test facility. In this paper, in order to verify reliability of DBE environment simulation test facility, statistics method is used

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)

DHS small-scale safety and thermal testing of improvised explosives-comparison of testing performance

International Nuclear Information System (INIS)

Reynolds, J G; Hsu, P C; Sandstrom, M M; Brown, G W; Warner, K F; Phillips, J J; Shelley, T J; Reyes, J A

2014-01-01

One of the first steps in establishing safe handling procedures for explosives is small-scale safety and thermal (SSST) testing. To better understand the response of improvised materials or homemade explosives (HMEs) to SSST testing, 16 HME materials were compared to three standard military explosives in a proficiency-type round robin study among five laboratories-two DoD and three DOE-sponsored by DHS. The testing matrix has been designed to address problems encountered with improvised materials-powder mixtures, liquid suspensions, partially wetted solids, immiscible liquids, and reactive materials. More than 30 issues have been identified that indicate standard test methods may require modification when applied to HMEs to derive accurate sensitivity assessments needed for developing safe handling and storage practices. This paper presents a generalized comparison of the results among the testing participants, comparison of friction results from BAM (German Bundesanstalt für Materi-alprüfung) and ABL (Allegany Ballistics Laboratory) designed testing equipment, and an overview of the statistical results from the RDX (1,3,5-Trinitroperhydro-1,3,5-triazine) standard tested throughout the proficiency test.

Testing methods of steel wi re ropes at the anchor

Directory of Open Access Journals (Sweden)

Stanislav Kropuch

2012-12-01

Full Text Available The present paper introduces an application of the acoustic andthermographic method in the defectoscopic testing of immobilesteel wire ropes at the most critical point, the anchor. Firstmeasurements and their results by these new defectoscopic methodsare shown. In defectoscopic tests at the anchor, the widelyused magnetic method gives unreliable results, and therefore presentsa problem for steel wire defectoscopy. Application of the two new methods in the steel wire defectoscopy at the anchor point will enableincreased safety measures at the anchor of steel wire ropes in bridge, roof, tower and aerial cable lift constructions.

43 CFR 11.64 - Injury determination phase-testing and sampling methods.

Science.gov (United States)

2010-10-01

... Institute for Occupational Safety and Health, the American Society for Testing and Materials, and the American Public Health Association. (iii) Methods selected for air sampling and analysis shall be methods... stresses simulated; (B) How the model approximates the geohydrological framework of the assessment area; (C...

Standard test method for determining the susceptibility to intergranular corrosion of 5XXX series Aluminum alloys by mass loss after exposure to nitric acid (NAMLT Test)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2004-01-01

1.1 This test method describes a procedure for constant immersion intergranular corrosion testing of 5XXX series aluminum alloys. 1.2 This test method is applicable only to wrought products. 1.3 This test method covers type of specimen, specimen preparation, test environment, and method of exposure. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

T. A. Tomberlin; S. B. Grover

2004-11-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

T. A. Tomberlin; S. B. Grover

2004-01-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

Guide On Safety Tests

Energy Technology Data Exchange (ETDEWEB)

NONE

1987-05-15

This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

Guide On Safety Tests

International Nuclear Information System (INIS)

1987-05-01

This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

Safety Testing of Ammonium Nitrate Based Mixtures

Science.gov (United States)

Phillips, Jason; Lappo, Karmen; Phelan, James; Peterson, Nathan; Gilbert, Don

2013-06-01

Ammonium nitrate (AN)/ammonium nitrate based explosives have a lengthy documented history of use by adversaries in acts of terror. While historical research has been conducted on AN-based explosive mixtures, it has primarily focused on detonation performance while varying the oxygen balance between the oxidizer and fuel components. Similarly, historical safety data on these materials is often lacking in pertinent details such as specific fuel type, particle size parameters, oxidizer form, etc. A variety of AN-based fuel-oxidizer mixtures were tested for small-scale sensitivity in preparation for large-scale testing. Current efforts focus on maintaining a zero oxygen-balance (a stoichiometric ratio for active chemical participants) while varying factors such as charge geometry, oxidizer form, particle size, and inert diluent ratios. Small-scale safety testing was conducted on various mixtures and fuels. It was found that ESD sensitivity is significantly affected by particle size, while this is less so for impact and friction. Thermal testing is in progress to evaluate hazards that may be experienced during large-scale testing.

Instrumentation and measurement method for the ATLAS test facility

Energy Technology Data Exchange (ETDEWEB)

Yun, Byong Jo; Chu, In Chul; Eu, Dong Jin; Kang, Kyong Ho; Kim, Yeon Sik; Song, Chul Hwa; Baek, Won Pil

2007-03-15

An integral effect test loop for pressurized water reactors (PWRs), the ATLAS is constructed by thermal-hydraulic safety research division in KAERI. The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400 which is a Korean evolution type nuclear reactors. A total 1300 instrumentations is equipped in the ATLAS test facility. In this report, the instrumentation of ATLAS test facility and related measurement methods were introduced.

A Proposal for a Methodology to Develop a Cyber-Attack Penetration Test Scenario Including NPPs Safety

Energy Technology Data Exchange (ETDEWEB)

Lee, In Hyo [KAIST, Daejeon (Korea, Republic of); Son, Han Seong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Si Won [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of); Kang, Hyun Gook [Rensselaer Polytechnic Institute, Troy (United States)

2016-10-15

Penetration test is a method to evaluate the cyber security of NPPs; so, this approach was performed in some studies. Because they focused on vulnerability finding or test bed construction, scenario based approach was not performed. However, to test the cyber security of NPPs, a proper test scenario should be needed. Ahn et al. developed cyber-attack scenarios but those scenarios couldn't be applied in penetration test because they developed the scenarios based on past incidents of NPPs induced by cyber-attack. That is, those scenarios only covered scenarios which were happened before; so, they couldn't cover other various scenarios and couldn't reflect them into a penetration test. In this study, a method to develop a cyber-attack penetration test scenario of NPPs especially focused on safety point of view is suggested. To evaluate the cyber security of NPPs, penetration test can be a possible way. In this study, a method to develop a penetration test scenario was explained. Especially, the goal of hacker was focused on nuclear fuel integrity deterioration. So, in the methodology, Level 1 PSA results were utilized to reflect plant safety into the security. From the PSA results, basic event was post processed and possible cyber-attacks were reviewed with vulnerabilities of digital control system.

A Proposal for a Methodology to Develop a Cyber-Attack Penetration Test Scenario Including NPPs Safety

International Nuclear Information System (INIS)

Lee, In Hyo; Son, Han Seong; Kim, Si Won; Kang, Hyun Gook

2016-01-01

Penetration test is a method to evaluate the cyber security of NPPs; so, this approach was performed in some studies. Because they focused on vulnerability finding or test bed construction, scenario based approach was not performed. However, to test the cyber security of NPPs, a proper test scenario should be needed. Ahn et al. developed cyber-attack scenarios but those scenarios couldn't be applied in penetration test because they developed the scenarios based on past incidents of NPPs induced by cyber-attack. That is, those scenarios only covered scenarios which were happened before; so, they couldn't cover other various scenarios and couldn't reflect them into a penetration test. In this study, a method to develop a cyber-attack penetration test scenario of NPPs especially focused on safety point of view is suggested. To evaluate the cyber security of NPPs, penetration test can be a possible way. In this study, a method to develop a penetration test scenario was explained. Especially, the goal of hacker was focused on nuclear fuel integrity deterioration. So, in the methodology, Level 1 PSA results were utilized to reflect plant safety into the security. From the PSA results, basic event was post processed and possible cyber-attacks were reviewed with vulnerabilities of digital control system

Relevance of microbial finished product testing in food safety management

NARCIS (Netherlands)

Zwietering, M.H.; Jacxsens, L.; Membre, J.M.; Nauta, M.; Peterz, M.

2016-01-01

Management of microbiological food safety is largely based on good design of processes, products and procedures. Finished product testing may be considered as a control measure at the end of the production process. However, testing gives only very limited information on the safety status of a food.

The Safety Feature Test of QNX RTOS

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Lee, Young Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-05-15

Benchmarking is a point of reference by which something can be measured. The QNX is a kind of Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available PC to be applied such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS-AC). The ELMSYS PC Hardware will be qualified by KTL(Korea Testing Lab.) in order to use as a Cabinet Operator Module(COM). QNX RTOS is dedicating by KAERI now. This paper describes the outline and some safety features among benchmarking test for QNX RTOS under the ELMSYS PC platform

The Safety Feature Test of QNX RTOS

International Nuclear Information System (INIS)

Kim, Jang Yeol; Lee, Young Jun

2010-01-01

Benchmarking is a point of reference by which something can be measured. The QNX is a kind of Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available PC to be applied such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS-AC). The ELMSYS PC Hardware will be qualified by KTL(Korea Testing Lab.) in order to use as a Cabinet Operator Module(COM). QNX RTOS is dedicating by KAERI now. This paper describes the outline and some safety features among benchmarking test for QNX RTOS under the ELMSYS PC platform

Standard Reference Test Method for Making Potentiostatic and Potentiodynamic Anodic Polarization Measurements

CERN Document Server

American Society for Testing and Materials. Philadelphia

2004-01-01

1.1 This test method covers an experimental procedure for checking experimental technique and instrumentation. If followed, this test method will provide repeatable potentiostatic and potentiodynamic anodic polarization measurements that will reproduce data determined by others at other times and in other laboratories provided all laboratories are testing reference samples from the same lot of Type 430 stainless steel. 1.2 Values stated in SI units are to be regarded as the standard. Inch-pound units given in parentheses are for information only. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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...

Standard Test Method for Abrasive Wear Resistance of Cemented

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 This test method covers the determination of abrasive wear resistance of cemented carbides. 1.2 The values stated in inch-pound units are to be regarded as the standard. The SI equivalents of inch-pound units are in parentheses and may be approximate. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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.)

16 CFR 1213.4 - Test methods.

Science.gov (United States)

2010-01-01

... Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT REGULATIONS SAFETY... orientation”). Then gradually apply a 33-lbf (147-N) force in a direction perpendicular to the plane of the... Figure 2 of this part into the portion of the boundary of the opening to be tested, with the plane of the...

Stress Tests Worldwide - IAEA Nuclear Safety Action Plan

International Nuclear Information System (INIS)

Lyons, J.E.

2012-01-01

The IAEA nuclear safety action plan relies on 11 important issues. 1) Safety assessments in light of the Fukushima accident: the IAEA secretariat will develop a methodology for stress tests against specific extreme natural hazards and will provide assistance for their implementation; 2) Strengthen existing IAEA peer reviews; 3) Emergency preparedness and response; 4) National Regulatory bodies in terms of independence and adequacy of human and financial resources; 5) The development of safety culture and scientific and technical capacity in Operating Organizations; 6) The upgrading of IAEA safety standards in a more efficient way; 7) A better implementation of relevant conventions concerning nuclear safety and nuclear accidents; 8) To provide a broad assistance on safety standard for countries embarking on a nuclear power program; 9) To facilitate the use of available information, expertise and techniques concerning radiation protection; 10) To enhance the transparency of nuclear industry; and 11) To promote the cooperation between member states in nuclear safety. (A.C.)

Safety testing of GM-rice expressing PHA-E lectin using a new animal test design

DEFF Research Database (Denmark)

Poulsen, Morten; Schrøder, Malene; Wilcks, Andrea

2007-01-01

The 90-day animal study is the core study for the safety assessment of genetically modified foods in the SAFOTEST project. The model compound tested in the 90-day study was a rice variety expressing the kidney bean Phaseolus vulgaris lectin agglutinin E-form (PHA-E lectin). Female Wistar rats were...... safety testing of genetically modified foods....

Standard test method for compressive (crushing) strength of fired whiteware materials

CERN Document Server

American Society for Testing and Materials. Philadelphia

2006-01-01

1.1 This test method covers two test procedures (A and B) for the determination of the compressive strength of fired whiteware materials. 1.2 Procedure A is generally applicable to whiteware products of low- to moderately high-strength levels (up to 150 000 psi or 1030 MPa). 1.3 Procedure B is specifically devised for testing of high-strength ceramics (over 100 000 psi or 690 MPa). 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Methods for Solar Energy Transmittance and Reflectance (Terrestrial) of Sheet Materials

CERN Document Server

American Society for Testing and Materials. Philadelphia

1971-01-01

1.1 These test methods cover the measurement of solar energy transmittance and reflectance (terrestrial) of materials in sheet form. Method A, using a spectrophotometer, is applicable for both transmittance and reflectance and is the referee method. Method B is applicable only for measurement of transmittance using a pyranometer in an enclosure and the sun as the energy source. Specimens for Method A are limited in size by the geometry of the spectrophotometer while Method B requires a specimen 0.61 m2 (2 ft2). For the materials studied by the drafting task group, both test methods give essentially equivalent results. 1.2 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Tensile and burst tests in support of the cadmium safety rod failure evaluation

International Nuclear Information System (INIS)

Thomas, J.K.

1992-02-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

Standard test method for measurement of web/roller friction characteristics

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This test method covers the simulation of a roller/web transport tribosystem and the measurement of the static and kinetic coefficient of friction of the web/roller couple when sliding occurs between the two. The objective of this test method is to provide users with web/roller friction information that can be used for process control, design calculations, and for any other function where web/roller friction needs to be known. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

Standard Test Method for Dust Erosion Resistance of Optical and Infrared Transparent Materials and Coatings

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 This test method covers the resistance of transparent plastics and coatings used in aerospace windscreens, canopies, and viewports to surface erosion as a result of dust impingement. This test method simulates flight through a defined particle cloud environment by means of independent control of particle size, velocity, impact angle, mass loading, and test duration. 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

Safety assessment of botanicals and botanical preparations used as ingredients in food supplements: testing an European Food Safety Authority-tiered approach.

Science.gov (United States)

Speijers, Gerrit; Bottex, Bernard; Dusemund, Birgit; Lugasi, Andrea; Tóth, Jaroslav; Amberg-Müller, Judith; Galli, Corrado L; Silano, Vittorio; Rietjens, Ivonne M C M

2010-02-01

This article describes results obtained by testing the European Food Safety Authority-tiered guidance approach for safety assessment of botanicals and botanical preparations intended for use in food supplements. Main conclusions emerging are as follows. (i) Botanical ingredients must be identified by their scientific (binomial) name, in most cases down to the subspecies level or lower. (ii) Adequate characterization and description of the botanical parts and preparation methodology used is needed. Safety of a botanical ingredient cannot be assumed only relying on the long-term safe use of other preparations of the same botanical. (iii) Because of possible adulterations, misclassifications, replacements or falsifications, and restorations, establishment of adequate quality control is necessary. (iv) The strength of the evidence underlying concerns over a botanical ingredient should be included in the safety assessment. (v) The matrix effect should be taken into account in the safety assessment on a case-by-case basis. (vi) Adequate data and methods for appropriate exposure assessment are often missing. (vii) Safety regulations concerning toxic contaminants have to be complied with. The application of the guidance approach can result in the conclusion that safety can be presumed, that the botanical ingredient is of safety concern, or that further data are needed to assess safety.

Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

International Nuclear Information System (INIS)

Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

1992-01-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

Test study on safety features of station blackout accident for nuclear main pump

International Nuclear Information System (INIS)

Liu Xiajie; Wang Dezhong; Zhang Jige; Liu Junsheng; Yang Zhe

2009-01-01

The theoretical and experimental studies of reactor coolant pump accidents encountered nation-wide and world-wide were described. To investigate the transient hydrodynamic performance of reactor coolant pump (RCP) during the period of rotational inertia in the station blackout accident, some theoretical and experimental studies were carried out, and the analysis of the test results was presented. The experiment parameters, conditions and test methods were introduced. The flow-rate, rotate speed and vibrations were analyzed emphatically. The quadruplicate polynomial curve equation was used to simulate the flow-rate,rotate speed along with time. The test results indicate that the flow-rate and rotator speed decrease rapidly at the very beginning of cut power and the test results accord with the regulation of safety standard. The vibrant displacement of bearing seat is intensified at the moment of lose power, but after a certain period rotor shaft libration changes. The test and analysis results help to understand the hydrodynamic performance of nuclear primary pump under lost of power accident, and provide the basic reference for safety evaluation. (authors)

Comparison of testing methods for particulate filters

International Nuclear Information System (INIS)

Ullmann, W.; Przyborowski, S.

1983-01-01

Four testing methods for particulate filters were compared by using the test rigs of the National Board of Nuclear Safety and Radiation Protection: 1) Measurement of filter penetration P as a function of particle size d by using a polydisperse NaC1 test aerosol and a scintillation particle counter; 2) Modified sodium flame test for measurement of total filter penetration P for various polydisperse NaC1 test aerosols; 3) Measurement of total filter penetration P for a polydisperse NaC1 test aerosol labelled with short-lived radon daughter products; 4) Measurement of total filter penetration P for a special paraffin oil test aerosol (oil fog test used in FRG according DIN 24 184, test aerosol A). The investigations were carried out on sheets of glass fibre paper (five grades of paper). Detailed information about the four testing methods and the used particle size distributions is given. The different results of the various methods are the base for the discussion of the most important parameters which influence the filter penetration P. The course of the function P=f(d) shows the great influence of the particle size. As expected there was also found a great dependence both from the test aerosol as well as from the principle and the measuring range of the aerosol-measuring device. The differences between the results of the various test methods are greater the lower the penetration. The use of NaCl test aerosol with various particle size distributions gives great differences for the respective penetration values. On the basis of these results and the values given by Dorman conclusions are made about the investigation of particulate filters both for the determination of filter penetration P as well as for the leak test of installed filters

Development and Testing of a Nutrition, Food Safety, and Physical Activity Checklist for EFNEP and FSNE Adult Programs

Science.gov (United States)

Bradford, Traliece; Serrano, Elena L.; Cox, Ruby H.; Lambur, Michael

2010-01-01

Objective: To develop and assess reliability and validity of the Nutrition, Food Safety, and Physical Activity Checklist to measure nutrition, food safety, and physical activity practices among adult Expanded Food and Nutrition Education Program (EFNEP) and Food Stamp Nutrition Education program (FSNE) participants. Methods: Test-retest…

Gas reactor in-pile safety test project (GRIST-2)

International Nuclear Information System (INIS)

Kelley, A.P. Jr.; Arbtin, E.; St Pierre, R.

1979-01-01

Although out-of-pile tests may be expected to confirm individual phenomena models in core disruptive accident analysis codes, only in-pile tests are capable of verifying the extremely complex integrated model effects within the appropriate time phase for these accidents. For this reason, the GRIST-2 project, the purpose of which is to design and construct an in-pile helium loop capable of transient safety testing in the TREAT facility in Idaho, forms a cornerstone of the US GCFR safety program. The project organization, experiment program, facility, helium system design, and schedule which have been selected to meet the objectives are described

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).

78 FR 71036 - Pipeline Safety: Random Drug Testing Rate; Contractor Management Information System Reporting...

Science.gov (United States)

2013-11-27

... PHMSA-2013-0248] Pipeline Safety: Random Drug Testing Rate; Contractor Management Information System Reporting; and Obtaining Drug and Alcohol Management Information System Sign-In Information AGENCY: Pipeline... Management Information System (MIS) Data; and New Method for Operators to Obtain User Name and Password for...

Optimization of safety equipment outages improves safety

International Nuclear Information System (INIS)

Cepin, Marko

2002-01-01

Testing and maintenance activities of safety equipment in nuclear power plants are an important potential for risk and cost reduction. An optimization method is presented based on the simulated annealing algorithm. The method determines the optimal schedule of safety equipment outages due to testing and maintenance based on minimization of selected risk measure. The mean value of the selected time dependent risk measure represents the objective function of the optimization. The time dependent function of the selected risk measure is obtained from probabilistic safety assessment, i.e. the fault tree analysis at the system level and the fault tree/event tree analysis at the plant level, both extended with inclusion of time requirements. Results of several examples showed that it is possible to reduce risk by application of the proposed method. Because of large uncertainties in the probabilistic safety assessment, the most important result of the method may not be a selection of the most suitable schedule of safety equipment outages among those, which results in similarly low risk. But, it may be a prevention of such schedules of safety equipment outages, which result in high risk. Such finding increases the importance of evaluation speed versus the requirement of getting always the global optimum no matter if it is only slightly better that certain local one

Standard test method for initial screening of corrosion inhibiting admixtures for steel in concrete

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 This test method covers a procedure for determining the effects of chemical admixtures on the corrosion of metals in concrete. This test method can be used to evaluate materials intended to inhibit chloride-induced corrosion of steel in concrete. It can also be used to evaluate the corrosivity of admixtures by themselves or in a chloride environment. This test is not applicable for emulsions. 1.2 &solely-SI-units; 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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).

The anchors of steel wire ropes, testing methods and their results

Directory of Open Access Journals (Sweden)

J. Krešák

2012-10-01

Full Text Available The present paper introduces an application of the acoustic and thermographic method in the defectoscopic testing of immobile steel wire ropes at the most critical point, the anchor. First measurements and their results by these new defectoscopic methods are shown. In defectoscopic tests at the anchor, the widely used magnetic method gives unreliable results, and therefore presents a problem for steel wire defectoscopy. Application of the two new methods in the steel wire defectoscopy at the anchor point will enable increased safety measures at the anchor of steel wire ropes in bridge, roof, tower and aerial cable lift constructions.

Evaluating the Effectiveness of Traditional Training Methods in Non-Traditional Training Programs for Adult Learners through a Pre-Test/Post-Test Comparison of Food Safety Knowledge

Science.gov (United States)

Dodd, Caleb D.; Burris, Scott; Fraze, Steve; Doerfert, David; McCulloch, Abigail

2013-01-01

The incorporation of hot and cold food bars into grocery stores in an effort to capture a portion of the home meal replacement industry is presenting new challenges for retail food establishments. To ensure retail success and customer safety, employees need to be educated in food safety practices. Traditional methods of training are not meeting…

Integrated Data Collection Analysis (IDCA) Program - SSST Testing Methods

Energy Technology Data Exchange (ETDEWEB)

Sandstrom, Mary M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Geoffrey W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Preston, Daniel N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pollard, Colin J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Warner, Kirstin F. [Naval Surface Warfare Center (NSWC), Indian Head, MD (United States). Indian Head Division; Remmers, Daniel L. [Naval Surface Warfare Center (NSWC), Indian Head, MD (United States). Indian Head Division; Sorensen, Daniel N. [Naval Surface Warfare Center (NSWC), Indian Head, MD (United States). Indian Head Division; Whinnery, LeRoy L. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Phillips, Jason J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Shelley, Timothy J. [Bureau of Alcohol, Tobacco and Firearms (ATF), Huntsville, AL (United States); Reyes, Jose A. [Applied Research Associates, Tyndall AFB, FL (United States); Hsu, Peter C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reynolds, John G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2013-03-25

The Integrated Data Collection Analysis (IDCA) program is conducting a proficiency study for Small- Scale Safety and Thermal (SSST) testing of homemade explosives (HMEs). Described here are the methods used for impact, friction, electrostatic discharge, and differential scanning calorimetry analysis during the IDCA program. These methods changed throughout the Proficiency Test and the reasons for these changes are documented in this report. The most significant modifications in standard testing methods are: 1) including one specified sandpaper in impact testing among all the participants, 2) diversifying liquid test methods for selected participants, and 3) including sealed sample holders for thermal testing by at least one participant. This effort, funded by the Department of Homeland Security (DHS), is putting the issues of safe handling of these materials in perspective with standard military explosives. The study is adding SSST testing results for a broad suite of different HMEs to the literature. Ultimately the study will suggest new guidelines and methods and possibly establish the SSST testing accuracies needed to develop safe handling practices for HMEs. Each participating testing laboratory uses identical test materials and preparation methods wherever possible. The testing performers involved are Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Indian Head Division, Naval Surface Warfare Center, (NSWC IHD), Sandia National Laboratories (SNL), and Air Force Research Laboratory (AFRL/RXQL). These tests are conducted as a proficiency study in order to establish some consistency in test protocols, procedures, and experiments and to compare results when these testing variables cannot be made consistent.

Standard test method for radiochemical determination of uranium isotopes in urine by alpha spectrometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 This test method is applicable to the determination of uranium in urine at levels of detection dependent on sample size, count time, detector background, and tracer yield. It is designed as a screening tool for detection of possible exposure of occupational workers. 1.2 This test method is designed for 50 mL of urine. This test method does not address the sampling protocol or sample preservation methods associated with its use. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

International Nuclear Information System (INIS)

Hayakawa, Masahiko

1981-01-01

The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

Construction of Earthquake - Proof Safety Evaluaiton Methods for Pipes with Wall Thinning

International Nuclear Information System (INIS)

Miyano, H.; Sekimura, N.; Takizawa, M.; Mastumoto, M.

2012-01-01

Since the Fukushima Dai-ichi accident, the importance of 'system safety' has been recognized anew. Particularly, system safety assessment of plants in operation from the various degradation perspectives, specifically, transition of time is very important. Accordingly, assessment on degradation will focus on the degradation of functions with passing of time, combined with the changes in the safety standards and concept of safety. Reliability assessment will be made on the consolidation of important functions, and not on individual components. The boundary function of the system will be one of the focus of this study. For the purpose of reliability assessment on the system by evaluating and quantifying the damage (or rupture) risk of piping - method for confirming the integrity of the system through the assessment on the damage (rupture) risk of the system when an external force caused by an earthquake is applied (the system is sound if the damage (rupture) risk is small) was examined on the basis of the prediction results for each of the parts in pipe wall thinning. In the next phase, the prediction results will be verified by tests, whereby, the improvement in reliability will be confirmed, and a combined assessment will be made in relation to the degradation factors of other systems. 'System safety' assessment method of plants in operation will be developed in a manner where a comprehensive assessment on the safety of the entire plant can be made. Specifically, the changes in the conditions, such as material degradations that degrade performance will be assessed on the entire system. Whereby, the risk caused by functional failure (damage) due to degradation will be regarded as the total of risk in the assessment. A framework on safety assessment will be structured, where the degree of safety will be measured by functional degradation, taking into consideration the changes made in the safety standards up to present. (author)

Allowed outage time for test and maintenance - Optimization of safety

International Nuclear Information System (INIS)

Cepin, M.; Mavko, B.

1997-01-01

The main objective of the project is the development and application of methodologies for improvement and optimization of test and maintenance activities for safety related equipment in NPPs on basis of their enhanced safety. The probabilistic safety assessment serves as a base, which does not mean the replacement of the deterministic analyses but the consideration of probabilistic safety assessment results as complement to deterministic results. 15 refs, 2 figs

Safer Roads: Comparisons Between Road Assessment Program and Composite Road Safety Index Method

Directory of Open Access Journals (Sweden)

Mohd Razelan Intan Suhana

2017-01-01

Full Text Available In most countries, crash statistics have becoming very crucial in evaluating road’s safety level. In Malaysia, these data are very important in deciding crash-prone areas known as black spot where specific road improvements plan will be proposed. However due to the unavailability of reliable crash data in many developing countries, appropriate road maintenance measures are facing great troubles. In light of that, several proactive methods in defining road’s safety level such as Road Assessment Program (RAP have emerged. This research aim to compare two proactive methods that have been tested in Malaysian roads ; road assessment program and road environment risk index which was developed based on composite index theory in defining road’s safety level. Composite road environment risk index was combining several crucial environment indicators, assigning weight and aggregating the individual index together to form a single value representing the road’s safety level. Based on the results, it can be concluded that both road assessment program and composite road environment risk index are contradicted in six different ways such as type of speed used, type of analysis used and their final outcomes. However, with an aim to promote safer roads, these two methods can be used concurrently as the outcomes in both methods seems to fulfil each other’s gap very well.

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.

Testing existing software for safety-related applications. Revision 7.1

International Nuclear Information System (INIS)

Scott, J.A.; Lawrence, J.D.

1995-12-01

The increasing use of commercial off-the-shelf (COTS) software products in digital safety-critical applications is raising concerns about the safety, reliability, and quality of these products. One of the factors involved in addressing these concerns is product testing. A tester's knowledge of the software product will vary, depending on the information available from the product vendor. In some cases, complete source listings, program structures, and other information from the software development may be available. In other cases, only the complete hardware/software package may exist, with the tester having no knowledge of the internal structure of the software. The type of testing that can be used will depend on the information available to the tester. This report describes six different types of testing, which differ in the information used to create the tests, the results that may be obtained, and the limitations of the test types. An Annex contains background information on types of faults encountered in testing, and a Glossary of pertinent terms is also included. This study is pertinent for safety-related software at reactors

Chemical Safety Assessment Using Read-Across: Assessing the Use of Novel Testing Methods to Strengthen the Evidence Base for Decision Making

Science.gov (United States)

Amcoff, Patric; Benigni, Romualdo; Blackburn, Karen; Carney, Edward; Cronin, Mark; Deluyker, Hubert; Gautier, Francoise; Judson, Richard S.; Kass, Georges E.N.; Keller, Detlef; Knight, Derek; Lilienblum, Werner; Mahony, Catherine; Rusyn, Ivan; Schultz, Terry; Schwarz, Michael; Schüürmann, Gerrit; White, Andrew; Burton, Julien; Lostia, Alfonso M.; Munn, Sharon; Worth, Andrew

2015-01-01

Background Safety assessment for repeated dose toxicity is one of the largest challenges in the process to replace animal testing. This is also one of the proof of concept ambitions of SEURAT-1, the largest ever European Union research initiative on alternative testing, co-funded by the European Commission and Cosmetics Europe. This review is based on the discussion and outcome of a workshop organized on initiative of the SEURAT-1 consortium joined by a group of international experts with complementary knowledge to further develop traditional read-across and include new approach data. Objectives The aim of the suggested strategy for chemical read-across is to show how a traditional read-across based on structural similarities between source and target substance can be strengthened with additional evidence from new approach data—for example, information from in vitro molecular screening, “-omics” assays and computational models—to reach regulatory acceptance. Methods We identified four read-across scenarios that cover typical human health assessment situations. For each such decision context, we suggested several chemical groups as examples to prove when read-across between group members is possible, considering both chemical and biological similarities. Conclusions We agreed to carry out the complete read-across exercise for at least one chemical category per read-across scenario in the context of SEURAT-1, and the results of this exercise will be completed and presented by the end of the research initiative in December 2015. Citation Berggren E, Amcoff P, Benigni R, Blackburn K, Carney E, Cronin M, Deluyker H, Gautier F, Judson RS, Kass GE, Keller D, Knight D, Lilienblum W, Mahony C, Rusyn I, Schultz T, Schwarz M, Schüürmann G, White A, Burton J, Lostia AM, Munn S, Worth A. 2015. Chemical safety assessment using read-across: assessing the use of novel testing methods to strengthen the evidence base for decision making. Environ Health Perspect 123:1232�

The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

International Nuclear Information System (INIS)

Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

1987-03-01

A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

Standard test method for color and color difference of whitewares by abriged spectrophotometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2001-01-01

1.1 This test method describes the instrumental measurement of the reflection properties and color of ceramic glazes and other whitewares by the use of a spectrophotometer or spectrocolorimeter with a hemispherical optical measuring system, such as an integrating sphere. 1.2 The test method is suitable for use with most specimens having an exterior flat surface large enough to cover the spectrophotometer sample port. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Method for Water Absorption of Core Materials for Structural Sandwich Constructions

CERN Document Server

American Society for Testing and Materials. Philadelphia

2001-01-01

1.1 This test method covers the determination of the relative amount of water absorption by various types of structural core materials when immersed or in a high relative humidity environment. This test method is intended to apply to only structural core materials; honeycomb, foam, and balsa wood. 1.2 The values stated in SI units are to be regarded as the standard. The inch-pound units given may be approximate. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

Combustion Safety Simplified Test Protocol Field Study

Energy Technology Data Exchange (ETDEWEB)

Brand, L. [Gas Technology Inst., Des Plaines, IL (United States); Cautley, D. [Gas Technology Inst., Des Plaines, IL (United States); Bohac, D. [Gas Technology Inst., Des Plaines, IL (United States); Francisco, P. [Gas Technology Inst., Des Plaines, IL (United States); Shen, L. [Gas Technology Inst., Des Plaines, IL (United States); Gloss, S. [Gas Technology Inst., Des Plaines, IL (United States)

2015-11-01

Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives. A survey of state weatherization agencies on combustion safety issues, details of a field data collection instrumentation package, summary of data collected over seven months, data analysis and results are included. The project team collected field data on 11 houses in 2015.

A comparative study of formal methods for safety critical software in nuclear power plant

International Nuclear Information System (INIS)

Sohn, Se Do; Seong Poong Hyun

2000-01-01

The requirement of ultra high reliability of the safety critical software can not be demonstrated by testing alone. The specification based on formal method is recommended for safety system software. But there exist various kinds of formal methods, and this variety of formal method is recognized as an obstacle to the wide use of formal method. In this paper six different formal method have been applied to the same part of the functional requirements that is calculation algorithm intensive. The specification results were compared against the criteria that is derived from the characteristics that good software requirements specifications should have and regulatory body recommends to have. The application experience shows that the critical characteristics should be defined first, then appropriate method has to e selected. In our case, the Software Cost Reduction method was recommended for internal condition or calculation algorithm checking, and state chart method is recommended for the external behavioral description. (author)

A Method to Analyze Threats and Vulnerabilities by Using a Cyber Security Test-bed of an Operating NPP

International Nuclear Information System (INIS)

Kim, Yong Sik; Son, Choul Woong; Lee, Soo Ill

2016-01-01

In order to implement cyber security controls for an Operating NPP, a security assessment should conduct in advance, and it is essential to analyze threats and vulnerabilities for a cyber security risk assessment phase. It might be impossible to perform a penetration test or scanning for a vulnerability analysis because the test may cause adverse effects on the inherent functions of ones. This is the reason why we develop and construct a cyber security test-bed instead of using real I and C systems in the operating NPP. In this paper, we propose a method to analyze threats and vulnerabilities of a specific target system by using a cyber security test-bed. The test-bed is being developed considering essential functions of the selected safety and non-safety system. This paper shows the method to analyze threats and vulnerabilities of a specific target system by using a cyber security test-bed. In order to develop the cyber security test-bed with both safety and non-safety functions, test-bed functions analysis and preliminary threats and vulnerabilities identification have been conducted. We will determine the attack scenarios and conduct the test-bed based vulnerability analysis

A Method to Analyze Threats and Vulnerabilities by Using a Cyber Security Test-bed of an Operating NPP

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Sik; Son, Choul Woong; Lee, Soo Ill [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

In order to implement cyber security controls for an Operating NPP, a security assessment should conduct in advance, and it is essential to analyze threats and vulnerabilities for a cyber security risk assessment phase. It might be impossible to perform a penetration test or scanning for a vulnerability analysis because the test may cause adverse effects on the inherent functions of ones. This is the reason why we develop and construct a cyber security test-bed instead of using real I and C systems in the operating NPP. In this paper, we propose a method to analyze threats and vulnerabilities of a specific target system by using a cyber security test-bed. The test-bed is being developed considering essential functions of the selected safety and non-safety system. This paper shows the method to analyze threats and vulnerabilities of a specific target system by using a cyber security test-bed. In order to develop the cyber security test-bed with both safety and non-safety functions, test-bed functions analysis and preliminary threats and vulnerabilities identification have been conducted. We will determine the attack scenarios and conduct the test-bed based vulnerability analysis.

Some Findings from Thermal-Hydraulic Validation Tests for SMART Passive Safety System

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Sik; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Shin, Yong-Cheol; Min, Kyoung-Ho; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To satisfy the domestic and international needs for nuclear safety improvement after the Fukushima accident, an effort to improve its safety has been studied, and a Passive Safety System (PSS) for SMART has been designed. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMARTITL. Recently, a test program to validate the performance of the SMART PSS was launched and its scaled-down test facility was additionally installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests for the SMART PSS will be summarized. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide base data during the application phase of SDA revision and construction licensing. A test program to validate the performance of SMARS PSS was launched with an additional scaleddown test facility of SMART PSS, which will be installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests of the SMART passive safety system during 2013-2014 were summarized. They include a couple of SMART PSS tests using active pumps and several 1-train SMART PSS tests. From the test results it was estimated that the SMART PSS has sufficient cooling capability to deal with the SBLOCA scenario of SMART. During the SBLOCA scenario, in the CMT the water layer inventory was well stratified thermally and the safety injection water was injected efficiently into the RPV from the initial period and cools down the RCS properly.

Physics design of fast reactor safety test facilities for in-pile experiments

International Nuclear Information System (INIS)

Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

1976-01-01

A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

Combustion Safety Simplified Test Protocol Field Study

Energy Technology Data Exchange (ETDEWEB)

Brand, L [Gas Technology Inst., Des Plaines, IL (United States); Cautley, D. [Gas Technology Inst., Des Plaines, IL (United States); Bohac, D. [Gas Technology Inst., Des Plaines, IL (United States); Francisco, P. [Gas Technology Inst., Des Plaines, IL (United States); Shen, L. [Gas Technology Inst., Des Plaines, IL (United States); Gloss, S. [Gas Technology Inst., Des Plaines, IL (United States)

2015-11-05

"9Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives. A survey of state weatherization agencies on combustion safety issues, details of a field data collection instrumentation package, summary of data collected over seven months, data analysis and results are included. The project provides several key results. State weatherization agencies do not generally track combustion safety failures, the data from those that do suggest that there is little actual evidence that combustion safety failures due to spillage from non-dryer exhaust are common and that only a very small number of homes are subject to the failures. The project team collected field data on 11 houses in 2015. Of these homes, two houses that demonstrated prolonged and excessive spillage were also the only two with venting systems out of compliance with the National Fuel Gas Code. The remaining homes experienced spillage that only occasionally extended beyond the first minute of operation. Combustion zone depressurization, outdoor temperature, and operation of individual fans all provide statistically significant predictors of spillage.

Methods and tools used at the IPSN for the safety assessment of critical software

International Nuclear Information System (INIS)

Regnier, P.; Henry, J.Y.

1998-01-01

A significant feature of EDF's latest 1400MWe ''N4'' generation of pressurized water reactor (PWR) is the extensive use of computerized instrumentation and control, including a fully digital system for the reactor protection function. For the safety assessment of the software driving the operation of this digital reactor protection called SPIN, IPSN has developed and implemented a set of methods and tools. Using the lessons learned from this experience, IPSN has worked at improving those methods and tools, mainly trying to make them more automatic to use, and has participated in an international assessment exercise to test some other methods and tools, either new products on the market or self-developed products. As a result of these works, this paper presents an up to date overview of the IPSN methods and tools used for the assessment of safety critical software. This assessment, which consists of an analysis of all the documentation associated with the technical specifications and of a representative set of functions, is usually carried out in five steps: (1) critical examination of the documents, (2) evaluation of the quality of the code, (3) determination of the critical software components, (4) development of test cases and choice of testing strategy, (5) dynamic analysis (consistency and robustness). This paper also presents methods and tools developed or implemented by IPSN in order to: evaluate the completeness and consistency of specification and design documents written in natural language; build a model and simulate specification or design items; evaluate the quality of the source code; carry out FMEA analysis; run the binary code and perform tests (CLAIRE); perform random or mutational tests. (author)

The influence of technical testing methods on perceptions of constructions

International Nuclear Information System (INIS)

Oestberg, Gustaf

2002-01-01

In principle, methods of testing the reliability and safety of constructions are designed with the aim of representing critical conditions for their use. Because of variations within technical systems, including the one for which a certain test was once developed, the validity of the method also varies. The adoption of a standardized test for different applications had to be based on the assumption that the test deals adequately with a real situation. As a consequence of this acceptance of a trustworthy relationship between test and reality, an imprint of the former on the latter occurs. This effect may even cause the perception of a reality to be defined by the corresponding test. Examples of such an influence may be found in the field of materials testing. Related phenomena can be dealt with by reference to current concepts of cognitive science and psychology of the perception of images

THE FLUORBOARD A STATISTICALLY BASED DASHBOARD METHOD FOR IMPROVING SAFETY

International Nuclear Information System (INIS)

PREVETTE, S.S.

2005-01-01

The FluorBoard is a statistically based dashboard method for improving safety. Fluor Hanford has achieved significant safety improvements--including more than a 80% reduction in OSHA cases per 200,000 hours, during its work at the US Department of Energy's Hanford Site in Washington state. The massive project on the former nuclear materials production site is considered one of the largest environmental cleanup projects in the world. Fluor Hanford's safety improvements were achieved by a committed partnering of workers, managers, and statistical methodology. Safety achievements at the site have been due to a systematic approach to safety. This includes excellent cooperation between the field workers, the safety professionals, and management through OSHA Voluntary Protection Program principles. Fluor corporate values are centered around safety, and safety excellence is important for every manager in every project. In addition, Fluor Hanford has utilized a rigorous approach to using its safety statistics, based upon Dr. Shewhart's control charts, and Dr. Deming's management and quality methods

Standard test method for linear thermal expansion of glaze frits and ceramic whiteware materials by the interferometric method

CERN Document Server

American Society for Testing and Materials. Philadelphia

1995-01-01

1.1 This test method covers the interferometric determination of linear thermal expansion of premelted glaze frits and fired ceramic whiteware materials at temperatures lower than 1000°C (1830°F). 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

Method of safety evaluation in nuclear power plants

International Nuclear Information System (INIS)

Kuraszkiewicz, P.; Zahn, P.

1988-01-01

A novel quantitative technique for evaluating safety of subsystems of nuclear power plants based on expert estimations is presented. It includes methods of mathematical psychology recognizing the effect of subjective factors in the expert estimates and, consequently, contributes to further objectification of evaluation. It may be applied to complementing probabilistic safety assessment. As a result of such evaluations a characteristic 'safety of nuclear power plants' is obtained. (author)

Standard Test Method for Normal Spectral Emittance at Elevated Temperatures

CERN Document Server

American Society for Testing and Materials. Philadelphia

1972-01-01

1.1 This test method describes a highly accurate technique for measuring the normal spectral emittance of electrically conducting materials or materials with electrically conducting substrates, in the temperature range from 600 to 1400 K, and at wavelengths from 1 to 35 μm. 1.2 The test method requires expensive equipment and rather elaborate precautions, but produces data that are accurate to within a few percent. It is suitable for research laboratories where the highest precision and accuracy are desired, but is not recommended for routine production or acceptance testing. However, because of its high accuracy this test method can be used as a referee method to be applied to production and acceptance testing in cases of dispute. 1.3 The values stated in SI units are to be regarded as the standard. The values in parentheses are for information only. 1.4 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 stan...

Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

International Nuclear Information System (INIS)

Han, Sang Min; Seong, Poong Hyun

2014-01-01

This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness

Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

Energy Technology Data Exchange (ETDEWEB)

Han, Sang Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

2014-08-15

This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness.

The need to optimize inservice testing and inspection to enhance safety

International Nuclear Information System (INIS)

Perry, J.A.

1996-01-01

Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view

The need to optimize inservice testing and inspection to enhance safety

Energy Technology Data Exchange (ETDEWEB)

Perry, J.A.

1996-12-01

Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view.

Modelling blood safety

NARCIS (Netherlands)

Janssen, M.P.

2010-01-01

This thesis describes the development and application of methods and models to support decision making on safety measures aimed at preventing the transmission of infections by blood donors. Safety measures refer to screening tests for blood donors, quarantine periods for blood plasma, or methods for

Comparative study of in-situ filter test methods

International Nuclear Information System (INIS)

Marshall, M.; Stevens, D.C.

1981-01-01

Available methods of testing high efficiency particulate aerosol (HEPA) filters in-situ have been reviewed. In order to understand the relationship between the results produced by different methods a selection has been compared. Various pieces of equipment for generating and detecting aerosols have been tested and their suitability assessed. Condensation-nuclei, DOP (di-octyl phthalate) and sodium-flame in-situ filter test methods have been studied, using the 500 cfm (9000 m 3 /h) filter test rig at Harwell and in the field. Both the sodium-flame and DOP methods measure the penetration through leaks and filter material. However the measured penetration through filtered leaks depends on the aerosol size distribution and the detection method. Condensation-nuclei test methods can only be used to measure unfiltered leaks since condensation nuclei have a very low penetration through filtered leaks. A combination of methods would enable filtered and unfiltered leaks to be measured. A condensation-nucleus counter using n-butyl alcohol as the working fluid has the advantage of being able to detect any particle up to 1 μm in diameter, including DOP, and so could be used for this purpose. A single-particle counter has not been satisfactory because of interference from particles leaking into systems under extract, particularly downstream of filters, and because the concentration of the input aerosol has to be severely limited. The sodium-flame method requires a skilled operator and may cause safety and corrosion problems. The DOP method using a total light scattering detector has so far been the most satisfactory. It is fairly easy to use, measures reasonably low values of penetration and gives rapid results. DOP has had no adverse effect on HEPA filters over a long series of tests

Standard Test Method for Bond Strength of Ceramic Tile to Portland Cement Paste

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method covers the determination of the ability of glazed ceramic wall tile, ceramic mosaic tile, quarry tile, and pavers to be bonded to portland cement paste. This test method includes both face-mounted and back-mounted tile. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Proof-testing strategies induced by dangerous detected failures of safety-instrumented systems

International Nuclear Information System (INIS)

Liu, Yiliu; Rausand, Marvin

2016-01-01

Some dangerous failures of safety-instrumented systems (SISs) are detected almost immediately by diagnostic self-testing as dangerous detected (DD) failures, whereas other dangerous failures can only be detected by proof-testing, and are therefore called dangerous undetected (DU) failures. Some items may have a DU- and a DD-failure at the same time. After the repair of a DD-failure is completed, the maintenance team has two options: to perform an insert proof test for DU-failure or not. If an insert proof test is performed, it is necessary to decide whether the next scheduled proof test should be postponed or performed at the scheduled time. This paper analyzes the effects of different testing strategies on the safety performance of a single channel of a SIS. The safety performance is analyzed by Petri nets and by approximation formulas and the results obtained by the two approaches are compared. It is shown that insert testing improves the safety performance of the channel, but the feasibility and cost of the strategy may be a hindrance to recommend insert testing. - Highlights: • Identify the tests induced by detected failures. • Model the testing strategies following DD-failures. • Propose analytical formulas for effects of strategies. • Simulate and verify the proposed models.

Recent metal fuel safety tests in TREAT

International Nuclear Information System (INIS)

Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

1986-01-01

In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

In-pile experiments and test facilities proposed for fast reactor safety

International Nuclear Information System (INIS)

Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

1976-01-01

The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

Standard test method for isotopic analysis of uranium hexafluoride by double standard single-collector gas mass spectrometer method

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This is a quantitative test method applicable to determining the mass percent of uranium isotopes in uranium hexafluoride (UF6) samples with 235U concentrations between 0.1 and 5.0 mass %. 1.2 This test method may be applicable for the entire range of 235U concentrations for which adequate standards are available. 1.3 This test method is for analysis by a gas magnetic sector mass spectrometer with a single collector using interpolation to determine the isotopic concentration of an unknown sample between two characterized UF6 standards. 1.4 This test method is to replace the existing test method currently published in Test Methods C761 and is used in the nuclear fuel cycle for UF6 isotopic analyses. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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 appro...

Research on fuzzy comprehensive assessment method of nuclear power plant safety culture

International Nuclear Information System (INIS)

Xiang Yuanyuan; Chen Xukun; Xu Rongbin

2012-01-01

Considering the traits of safety culture in nuclear plant, 38 safety culture assessment indexes are established from 4 aspects such as safety values, safety institution, safety behavior and safety sub- stances. Based on it, a comprehensive assessment method for nuclear power plant safety culture is constructed by using AHP (Analytic Hierarchy Process) approach and fuzzy mathematics. The comprehensive assessment method has the quality of high precision and high operability, which can support the decision making of safety culture development. (authors)

Validation of programmable industrial automation systems for safety critical applications in NPP's; dynamic testing

International Nuclear Information System (INIS)

Haapanen, P.; Korhonen, J.

1995-01-01

The safety assessment of programmable automation systems cannot be totally be based on conventional probabilistic methods because of the difficulties in quantification of the reliability of the software as well as the hardware. Additional means shall therefore be used to gain more confidence on the system dependability. One central confidence building measure is the independent dynamic testing of the completed system. An automated test harness is needed to run the required large amount of test cases in a restricted time span. This paper describes a prototype dynamic testing harness for programmable digital systems developed at VTT. (author). 12 refs, 2 figs, 2 tabs

Standard Test Method for Measuring Optical Angular Deviation of Transparent Parts

CERN Document Server

American Society for Testing and Materials. Philadelphia

1996-01-01

1.1 This test method covers measuring the angular deviation of a light ray imposed by transparent parts such as aircraft windscreens and canopies. The results are uncontaminated by the effects of lateral displacement, and the procedure may be performed in a relatively short optical path length. This is not intended as a referee standard. It is one convenient method for measuring angular deviations through transparent windows. 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

A Study on the Estimation Method of Risk Based Area for Jetty Safety Monitoring

Directory of Open Access Journals (Sweden)

Byeong-Wook Nam

2015-09-01

Full Text Available Recently, the importance of safety-monitoring systems was highlighted by the unprecedented collision between a ship and a jetty in Yeosu. Accordingly, in this study, we introduce the concept of risk based area and develop a methodology for a jetty safety-monitoring system. By calculating the risk based areas for a ship and a jetty, the risk of collision was evaluated. To calculate the risk based areas, we employed an automatic identification system for the ship, stopping-distance equations, and the regulation velocity near the jetty. In this paper, we suggest a risk calculation method for jetty safety monitoring that can determine the collision probability in real time and predict collisions using the amount of overlap between the two calculated risk based areas. A test was conducted at a jetty control center at GS Caltex, and the effectiveness of the proposed risk calculation method was verified. The method is currently applied to the jetty-monitoring system at GS Caltex in Yeosu for the prevention of collisions.

Emerging research methods and their application to road safety.

Science.gov (United States)

Tarko, Andrew; Boyle, Linda Ng; Montella, Alfonso

2013-12-01

The study of road safety has seen great strides over the past few decades with advances in analytical methods and research tools that allow researchers to provide insights into the complex interactions of the driver, vehicle, and roadway. Data collection methods range from traditional traffic and roadway sensors to instrumented vehicles and driving simulators, capable of providing detailed data on both the normal driving conditions and the circumstances surrounding a safety critical event. In September 2011, the Third International Conference on Road Safety and Simulation was held in Indianapolis, Indiana, USA, which was hosted by the Purdue University Center for Road Safety and sponsored by the Transportation Research Board and its three committees: ANB20 Safety Data, Analysis and Evaluation, AND30 Simulation and Measurement of Vehicle and Operator Performance, and ABJ95 Visualization in Transportation. The conference brought together two hundred researchers from all over the world demonstrating some of the latest research methods to quantify crash causality and associations, and model road safety. This special issue is a collection of 14 papers that were presented at the conference and then peer-reviewed through this journal. These papers showcase the types of analytical tools needed to examine various crash types, the use of naturalistic and on-road data to validate the use of surrogate measures of safety, and the value of driving simulators to examine high-risk situations. Copyright © 2013 Elsevier Ltd. All rights reserved.

Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

Energy Technology Data Exchange (ETDEWEB)

Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-09-01

An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

Classifying Secondary Task Driving Safety Using Method of F-ANP

Directory of Open Access Journals (Sweden)

Lisheng Jin

2015-02-01

Full Text Available This study was designed to build an evaluation system for secondary task driving safety by using method of Fuzzy Analytic Network Process (F-ANP. Forty drivers completed driving on driving simulator while interacting with or without a secondary task. Measures of fixations, saccades, and vehicle running status were analyzed. According to five experts' opinions, a hierarchical model for secondary task driving safety evaluation was built. The hierarchical model was divided into three levels: goal, assessment dimension, and criteria. Seven indexes make up the level of criteria, and the assessment dimension includes two clusters: vehicle control risk and driver eye movement risk. By method of F-ANP, the priorities of the criteria and the subcriteria were determined. Furthermore, to rank the driving safety, an approach based on the principle of maximum membership degree was adopted. At last, a case study of secondary task driving safety evaluation by forty drivers using the proposed method was done. The results indicated that the application of the proposed method is practically feasible and adoptable for secondary task driving safety evaluation.

A SOFTWARE RELIABILITY ESTIMATION METHOD TO NUCLEAR SAFETY SOFTWARE

Directory of Open Access Journals (Sweden)

GEE-YONG PARK

2014-02-01

Full Text Available A method for estimating software reliability for nuclear safety software is proposed in this paper. This method is based on the software reliability growth model (SRGM, where the behavior of software failure is assumed to follow a non-homogeneous Poisson process. Two types of modeling schemes based on a particular underlying method are proposed in order to more precisely estimate and predict the number of software defects based on very rare software failure data. The Bayesian statistical inference is employed to estimate the model parameters by incorporating software test cases as a covariate into the model. It was identified that these models are capable of reasonably estimating the remaining number of software defects which directly affects the reactor trip functions. The software reliability might be estimated from these modeling equations, and one approach of obtaining software reliability value is proposed in this paper.

A proactive method for safety management in nuclear facilities

International Nuclear Information System (INIS)

Grecco, Claudio Henrique dos Santos; Carvalho, Paulo Victor Rodrigues de; Santos, Isaac Antonio Luquetti dos

2014-01-01

Due to the modern approach to address the safety of nuclear facilities which highlights that these organizations must be able to assess and proactively manage their activities becomes increasingly important the need for instruments to evaluate working conditions. In this context, this work presents a proactive method of managing organizational safety, which has three innovative features: 1) the use of predictive indicators that provide current information on the performance of activities, allowing preventive actions and not just reactive in safety management, different from safety indicators traditionally used (reactive indicators) that are obtained after the occurrence of undesired events; 2) the adoption of resilience engineering approach in the development of indicators - indicators are based on six principles of resilience engineering: top management commitment, learning, flexibility, awareness, culture of justice and preparation for the problems; 3) the adoption of the concepts and properties of fuzzy set theory to deal with subjectivity and consistency of human trials in the evaluation of the indicators. The fuzzy theory is used primarily to map qualitative models of decision-making, and inaccurate representation methods. The results of this study aim an improvement in performance and safety in organizations. The method was applied in a radiopharmaceutical shipping sector of a nuclear facility. The results showed that the method is a good monitoring tool objectively and proactively of the working conditions of an organizational domain

Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

International Nuclear Information System (INIS)

Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

2013-01-01

Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

PC based manual and safety logic card test setup for 235 MWe PHWRs

International Nuclear Information System (INIS)

Chandgadkar, G.M.; Kohli, A.K.; Agarwal, R.G.; Chandra, Rajesh

1992-01-01

Fuel handling controls for 235 MWe PHWR make use of Manual and Logic cards (MLCs) for providing safety interlocks. These cards consist of various type of logic blocks. By connecting these logic blocks all the safety interlocks required for fuel handling controls have been provided. Previously trouble shooting of these cards was done by means of logic probe. Since the method was manual, it was laborious and time consuming. PC based test setup has overcome this drawback and detects the fault at the component level within few seconds. It also gives printout of status of faulty MLC cards. Here motherboard has been designed having slots for insertion of MLC cards. The input/output connection of these cards are coming to two 50 pin FRC connectors. PC communicates through 144 line digital input/output card with MLC card under test. Software is user friendly and outputs suitable input patterns to the card under test and checks for output pattern. It compares this output pattern with compare pattern and detects the fault and displays the symptoms. This system is currently in use at test facility for fuelling machine for 235 MWe PHWR reactor at Refuelling Technology Division, Hall-7. This test setup has been proposed for use at NAPP and future reactors. (author). 4 figs., 1 annexure

Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety

International Nuclear Information System (INIS)

Ammerman, D.J.

1997-01-01

Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A 2 quantities). These packages have the most stringent performance criteria. Material with less than an A 2 quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in open-quotes strong-tightclose quotes packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material

Standard test method for determination of surface lubrication on flexible webs

CERN Document Server

American Society for Testing and Materials. Philadelphia

1999-01-01

1.1 This test method has been used since 1988 as an ANSI/ISO standard test for determination of lubrication on processed photographic films. Its purpose was to determine the presence of process-surviving lubricants on photographic films. It is the purpose of this test method to expand the applicability of this test method to other flexible webs that may need lubrication for suitable performance. This test measures the breakaway (static) coefficient of friction of a metal rider on the web by the inclined plane method. The objectives of the test is to determine if a web surface has a lubricant present or not. It is not intended to assign a friction coefficient to a material. It is not intended to rank lubricants. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 ...

An Evaluation Method for Team Competencies to Enhance Nuclear Safety Culture

International Nuclear Information System (INIS)

Hang, S. M.; Seong, P. H.; Kim, A. R.

2016-01-01

Safety culture has received attention in safety-critical industries, including nuclear power plants (NPPs), due to various prominent accidents such as concealment of a Station Blackout (SBO) of Kori NPP unit 1 in 2012, the Sewol ferry accident in 2014, and the Chernobyl accident in 1986. Analysis reports have pointed out that one of the major contributors to the cause of the accidents is ‘the lack of safety culture’. The term, nuclear safety culture, was firstly defined after the Chernobyl accident by the IAEA in INSAG report no. 4, 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 their significance.” Afterwards, a wide consensus grew among researchers and nuclear-related organizations, that safety culture should be evaluated and managed in a certain manner. Consequently, each nuclear-related organization defined and developed their own safety culture definitions and assessment methods. However, none of these methods provides a way for an individual or a team to enhance the safety culture of an organization. Especially for a team, which is the smallest working unit in NPPs, team members easily overlook their required practices to improve nuclear safety culture. Therefore in this study, we suggested a method to estimate nuclear safety culture of a team, by approaching with the ‘competency’ point of view. The competency is commonly focused on individuals, and defined as, “underlying characteristics of an individual that are causally related to effective or superior performance in a job.” Similar to safety culture, the definition of competency focuses on characteristics and attitudes of individuals. Thus, we defined ‘safety culture competency’ as “underlying characteristics and outward attitudes of individuals that are causally related to a healthy and strong nuclear safety

Safety monitoring in process and control

International Nuclear Information System (INIS)

Esparza, V. Jr.; Sebo, D.E.

1984-01-01

Safety Functions provide a method of ensuring the safe operation of any large-scale processing plant. Successful implementation of safety functions requires continuous monitoring of safety function values and trends. Because the volume of information handled by a plant operator occassionally can become overwhelming, attention may be diverted from the primary concern of maintaining plant safety. With this in mind EG and G, Idaho developed various methods and techniques for use in a computerized Safety Function Monitoring System and tested the application of these techniques using a simulated nuclear power plant, the Loss-of-Fluid Test Facility (LOFT) at the Idaho National Engineering Laboratory (INEL). This paper presents the methods used in the development of a Safety Function Monitoring System

Safety and reliability of pressure components with special emphasis on advanced methods of NDT. Vol. 2

International Nuclear Information System (INIS)

1986-01-01

The 12 papers discuss topics of strength and safety in the field of materials technology and engineering. Conclusions for NPP component safety and materials are drawn. Measurements and studies relate to fracture mechanics methods (oscillation, burst, material strength, characteristics). The dynamic analysis of the behaviour of large test specimens, the influence of load velocity on crack resistance curve and the development of forged parts from austenitic steel for fast breeder reactors are presented. (DG) [de

Nondestructive testing method

International Nuclear Information System (INIS)

Porter, J.F.

1996-01-01

Nondestructive testing (NDT) is the use of physical and chemical methods for evaluating material integrity without impairing its intended usefulness or continuing service. Nondestructive tests are used by manufaturer's for the following reasons: 1) to ensure product reliability; 2) to prevent accidents and save human lives; 3) to aid in better product design; 4) to control manufacturing processes; and 5) to maintain a uniform quality level. Nondestructive testing is used extensively on power plants, oil and chemical refineries, offshore oil rigs and pipeline (NDT can even be conducted underwater), welds on tanks, boilers, pressure vessels and heat exchengers. NDT is now being used for testing concrete and composite materials. Because of the criticality of its application, NDT should be performed and the results evaluated by qualified personnel. There are five basic nondestructive examination methods: 1) liquid penetrant testing - method used for detecting surface flaws in materials. This method can be used for metallic and nonmetallic materials, portable and relatively inexpensive. 2) magnetic particle testing - method used to detect surface and subsurface flaws in ferromagnetic materials; 3) radiographic testing - method used to detect internal flaws and significant variation in material composition and thickness; 4) ultrasonic testing - method used to detect internal and external flaws in materials. This method uses ultrasonics to measure thickness of a material or to examine the internal structure for discontinuities. 5) eddy current testing - method used to detect surface and subsurface flaws in conductive materials. Not one nondestructive examination method can find all discontinuities in all of the materials capable of being tested. The most important consideration is for the specifier of the test to be familiar with the test method and its applicability to the type and geometry of the material and the flaws to be detected

Cyber Security Penetration Test for Digital Safety I and C Systems

International Nuclear Information System (INIS)

Lee, C. K.; Kim, D. H.; Kwon, K. C.; Joo, H. K.; Song, J. S.

2010-01-01

In the Korea Nuclear I and C Systems Development project the platforms for plant protection systems are developed, which function as a reactor shutdown, actuation of engineered safety features and a control of the related equipment. Those are fully digitalized through the use of safety-grade programmable logic controllers (PLCs) and few types of communication network. However the Regulatory Guide 1.152 (Rev. 02) was published by the U.S. NRC in 2006 and it recommended the application of a cyber security to the safety systems in the Nuclear Power Plant (NPP). Therefore to incorporate the new licensing requirement, a cyber security risk assessment is performed for the platforms. Then the vulnerabilities identified by the risk assessment are validated by penetration test. This paper summarizes test scenario, test results and their incorporation into system design

Probabilistic safety analysis : a new nuclear power plants licensing method

International Nuclear Information System (INIS)

Oliveira, L.F.S. de.

1982-04-01

After a brief retrospect of the application of Probabilistic Safety Analysis in the nuclear field, the basic differences between the deterministic licensing method, currently in use, and the probabilistic method are explained. Next, the two main proposals (by the AIF and the ACRS) concerning the establishment of the so-called quantitative safety goals (or simply 'safety goals') are separately presented and afterwards compared in their most fundamental aspects. Finally, some recent applications and future possibilities are discussed. (Author) [pt

Standard Test Methods for Measurement of Electrical Performance and Spectral Response of Nonconcentrator Multijunction Photovoltaic Cells and Modules

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 These test methods provide special techniques needed to determine the electrical performance and spectral response of two-terminal, multijunction photovoltaic (PV) devices, both cell and modules. 1.2 These test methods are modifications and extensions of the procedures for single-junction devices defined by Test Methods E948, E1021, and E1036. 1.3 These test methods do not include temperature and irradiance corrections for spectral response and current-voltage (I-V) measurements. Procedures for such corrections are available in Test Methods E948, E1021, and E1036. 1.4 These test methods may be applied to cells and modules intended for concentrator applications. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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 appropriate safety and health practices and ...

Environmental safety evaluation in test sea disposal of low-level radioactive wastes

International Nuclear Information System (INIS)

1979-01-01

The study results on the environmental safety in the test sea disposal of low-level wastes by Subcommittee on Radioactive Waste Safety Technology in Nuclear Safety Commission are given in connection with the test disposal of radioactive wastes into sea reported by the Nuclear Safety Bureau. The Subcommittee concludes that the effect of the test disposal of radioactive wastes into sea on the environment is extremely small. The contents are as follows. The full text of the report; attached data, (1) prediction of the concentrations of radioactive nuclides in sea, (2) calculation of the concentrations of radioactive nuclides in marine life with biological paths, and (3) estimation of exposure dose in general people; references (1) radiation exposure of the personnel engaged in sea disposal, (2) the effect of a sea disaster during ocean transport. (J.P.N.)

Validation of programmable industrial automation systems for safety critical applications in NPP's dynamic testing

International Nuclear Information System (INIS)

Haapanen, P.; Korhonen, J.

1995-01-01

The safety assessment of programmable automation systems can not totally be based on conventional probabilistic methods because of the difficulties in quantification of the reliability of the software as well as the hardware. Additional means shall therefore be used to gain more confidence on the system dependability. One central confidence building measure is the independent dynamic testing of the completed system. An automated test harness is needed to run the required large amount of test cases in a restricted time span. The prototype dynamic testing harness for programmable digital systems developed at the Technical Research Centre of Finland (VTT) is described in the presentation. (12 refs., 2 figs., 2 tabs.)

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

2005-03-01

Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

Standard Test Method for Measuring Heat-Transfer Rate Using a Thermal Capacitance (Slug) Calorimeter

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method describes the measurement of heat transfer rate using a thermal capacitance-type calorimeter which assumes one-dimensional heat conduction into a cylindrical piece of material (slug) with known physical properties. 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. Note 1—For information see Test Methods E 285, E 422, E 458, E 459, and E 511.

An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

International Nuclear Information System (INIS)

Waltar, A.E.; Padilla, A. Jr.

1990-11-01

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

Safety properties test data selection from an algebraic model of Lustre programs; Selection de tests de proprietes de surete a partir d'une modelisation algebrique de programmes lustre

Energy Technology Data Exchange (ETDEWEB)

Hsiao, N.Ch

1995-07-01

In the context to validate an industrial software, which is a set of reactive programs, we are confronted with the safety properties verification problem. This thesis reports an experience in which our goal is to generate the test data satisfying a safety property. The software to be validated is designed with the SAGA tool, in which a view can be regarded as a program of a programming language called LUSTRE. We adapt a test data generation tool called LOFT to this kind of programs. In this way, we consider the functional testing method on which the LOFT tool is based. Given any LUSTRE program, we try to give it an algebraic model because LOFT treats algebraic specifications. So, our task consists In defining a formal framework in which any LUSTRE program can be translated into a LOFT module: based on an operational semantics of the LUSTRE language, the flow types 'T-flow' are specified with the constructive algebraic formalism, then implemented in a LOFT modules base. Next, in a test selection process assisted by LOFT, a safety property Is expressed by an equation to join other control hypotheses, and to guide the test data selection. Some concrete test data set are generated in this way on some significant examples. This experience confirm the feasibility of formal method on test data selection for the reactive programs. (author)

Recommendations for safety testing with the in vivo comet assay.

Science.gov (United States)

Vasquez, Marie Z

2012-08-30

While the in vivo comet assay increases its role in regulatory safety testing, deliberations about the interpretation of comet data continue. Concerns can arise regarding comet assay publications with limited data from non-blind testing of positive control compounds and using protocols (e.g. dose concentrations, sample times, and tissues) known to give an expected effect. There may be a tendency towards bias when the validation or interpretation of comet assay data is based on results generated by widely accepted but non-validated assays. The greatest advantages of the comet assay are its sensitivity and its ability to detect genotoxicity in tissues and at sample times that could not previously be evaluated. Guidelines for its use and interpretation in safety testing should take these factors into account. Guidelines should be derived from objective review of data generated by blind testing of unknown compounds dosed at non-toxic concentrations and evaluated in a true safety-testing environment, where the experimental design and conclusions must be defensible. However, positive in vivo comet findings with such compounds are rarely submitted to regulatory agencies and this data is typically unavailable for publication due to its proprietary nature. To enhance the development of guidelines for safety testing with the comet assay, and with the permission of several sponsors, this paper presents and discusses relevant data from multiple GLP comet studies conducted blind, with unknown pharmaceuticals and consumer products. Based on these data and the lessons we have learned through the course of conducting these studies, I suggest significant adjustments to the current conventions, and I provide recommendations for interpreting in vivo comet assay results in situations where risk must be evaluated in the absence of carcinogenicity or clinical data. Copyright © 2012 Elsevier B.V. All rights reserved.

Suggestions on the Development of Safety Culture Assessment Method

International Nuclear Information System (INIS)

Choi, Young Sung; Choi, Kwang Sik; Kim, Woong Sik

2006-01-01

Several efforts have been made to assess safety culture of organization that operates nuclear power plants in Korea. The MOST and KINS played a major role to develop assessment methods and KHNP applied them to its NPPs. This paper explains the two methods developed by KINS briefly and presents the insights obtained from the two different applications. It concludes with some suggestions for safety culture assessment based on the insights

Crane Safety Assessment Method Based on Entropy and Cumulative Prospect Theory

Directory of Open Access Journals (Sweden)

Aihua Li

2017-01-01

Full Text Available Assessing the safety status of cranes is an important problem. To overcome the inaccuracies and misjudgments in such assessments, this work describes a safety assessment method for cranes that combines entropy and cumulative prospect theory. Firstly, the proposed method transforms the set of evaluation indices into an evaluation vector. Secondly, a decision matrix is then constructed from the evaluation vectors and evaluation standards, and an entropy-based technique is applied to calculate the index weights. Thirdly, positive and negative prospect value matrices are established from reference points based on the positive and negative ideal solutions. Thus, this enables the crane safety grade to be determined according to the ranked comprehensive prospect values. Finally, the safety status of four general overhead traveling crane samples is evaluated to verify the rationality and feasibility of the proposed method. The results demonstrate that the method described in this paper can precisely and reasonably reflect the safety status of a crane.

Safety demonstration test on solvent fire in fuel reprocessing plant

International Nuclear Information System (INIS)

Nishio, Gunji; Hashimoto, Kazuichiro

1989-03-01

This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

Safety test facilities. Needs and concepts. A French evaluation

International Nuclear Information System (INIS)

Tretiakoff, O.; Bailly, J.

1976-01-01

The fuel behavior of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program. These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The purpose of this paper is to explain how a logical set of simple observations has led us to the present state of the Cadarache in-pile experimental safety program and how it may help us to find our way in a dense forest of both technical and psychological difficulties

Performance Test Results of Safety I and C Systems of SMART MMIS

International Nuclear Information System (INIS)

Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong

2011-01-01

KAERI has developed SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, since 1997 and submitted a SSAR (Standard design Safety Analysis Report) to Korea institute of nuclear safety (KINS) at the end of 2010 for the purpose of achieving the standard design approval (SDA) by the end of 2011. SMART MMIS has been designed with fully digitalized systems. Non-safety instrumentation and control (I and C) systems are designed based on the commercial distributed control systems. The safety I and C systems are designed using a new platform that was developed and validated by KAERI. Safety I and C systems are modularized using the platform. In the protection systems (PSs), datalinks are used to transmit data in a one-way direction in order to meet the independency requirement. In the engineered safety features-component control system (ESF-CCS), network switch devices (NSDs) are used to connect the group and loop controllers. The NSD was also newly developed and validated by KAERI. After validating the platform and NSD, a test facility was developed using the platform and NSDs to validate the performance of safety I and C systems. This paper presents the development and test results from the test facility

Cyber Security Test Strategy for Non-safety Display System

International Nuclear Information System (INIS)

Son, Han Seong; Kim, Hee Eun

2016-01-01

Cyber security has been a big issue since the instrumentation and control (I and C) system of nuclear power plant (NPP) is digitalized. A cyber-attack on NPP should be dealt with seriously because it might cause not only economic loss but also the radioactive material release. Researches on the consequences of cyber-attack onto NPP from a safety point of view have been conducted. A previous study shows the risk effect brought by initiation of event and deterioration of mitigation function by cyber terror. Although this study made conservative assumptions and simplifications, it gives an insight on the effect of cyber-attack. Another study shows that the error on a non-safety display system could cause wrong actions of operators. According to this previous study, the failure of the operator action caused by a cyber-attack on a display system might threaten the safety of the NPP by limiting appropriate mitigation actions. This study suggests a test strategy focusing on the cyber-attack on the information and display system, which might cause the failure of operator. The test strategy can be suggested to evaluate and complement security measures. Identifying whether a cyber-attack on the information and display system can affect the mitigation actions of operator, the strategy to obtain test scenarios is suggested. The failure of mitigation scenario is identified first. Then, for the test target in the scenario, software failure modes are applied to identify realistic failure scenarios. Testing should be performed for those scenarios to confirm the integrity of data and to assure effectiveness of security measures

Cyber Security Test Strategy for Non-safety Display System

Energy Technology Data Exchange (ETDEWEB)

Son, Han Seong [Joongbu University, Geumsan (Korea, Republic of); Kim, Hee Eun [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Cyber security has been a big issue since the instrumentation and control (I and C) system of nuclear power plant (NPP) is digitalized. A cyber-attack on NPP should be dealt with seriously because it might cause not only economic loss but also the radioactive material release. Researches on the consequences of cyber-attack onto NPP from a safety point of view have been conducted. A previous study shows the risk effect brought by initiation of event and deterioration of mitigation function by cyber terror. Although this study made conservative assumptions and simplifications, it gives an insight on the effect of cyber-attack. Another study shows that the error on a non-safety display system could cause wrong actions of operators. According to this previous study, the failure of the operator action caused by a cyber-attack on a display system might threaten the safety of the NPP by limiting appropriate mitigation actions. This study suggests a test strategy focusing on the cyber-attack on the information and display system, which might cause the failure of operator. The test strategy can be suggested to evaluate and complement security measures. Identifying whether a cyber-attack on the information and display system can affect the mitigation actions of operator, the strategy to obtain test scenarios is suggested. The failure of mitigation scenario is identified first. Then, for the test target in the scenario, software failure modes are applied to identify realistic failure scenarios. Testing should be performed for those scenarios to confirm the integrity of data and to assure effectiveness of security measures.

Safety evaluation report. Fast Flux Test Facility. Project No. 448

Energy Technology Data Exchange (ETDEWEB)

1978-08-01

Information on the safety of the FFTF Reactor is presented under the following chapter headings: site characteristics; design of structures, components, equipment, and systems; reactor; reactor coolant system and connected systems; engineered safety features; electric power; auxiliary systems; radioactive waste management systems; radiation protection; conduct of operations; initial test programs; accident analysis; and quality assurance.

Safety evaluation report. Fast Flux Test Facility. Project No. 448

International Nuclear Information System (INIS)

1978-01-01

Information on the safety of the FFTF Reactor is presented under the following chapter headings: site characteristics; design of structures, components, equipment, and systems; reactor; reactor coolant system and connected systems; engineered safety features; electric power; auxiliary systems; radioactive waste management systems; radiation protection; conduct of operations; initial test programs; accident analysis; and quality assurance

Safety tests file

International Nuclear Information System (INIS)

2011-01-01

The design and operation of nuclear power plants is governed by strict and clearly defined regulations designed to ensure their safety in all circumstances. Since the first nuclear reactors were commissioned, the basic safety principles and the corresponding practical requirements have constantly evolved and been enhanced, benefiting from operating experience feedback from reactors around the world (about 500 production reactors currently in service). Reactor safety has from the outset been built around the 'defense in depth' concept, which aims to prevent melting of the core and radioactive releases into the environment. It can be summarized as follows: over and above all the measures taken to prevent accidents, the principle that accidents do occur has to be accepted. We then assess their consequences and take steps to contain them at the level of severity at which they occur. (authors)

Standard test method for quantitative determination of americium 241 in plutonium by Gamma-Ray spectrometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

1994-01-01

1.1 This test method covers the quantitative determination of americium 241 by gamma-ray spectrometry in plutonium nitrate solution samples that do not contain significant amounts of radioactive fission products or other high specific activity gamma-ray emitters. 1.2 This test method can be used to determine the americium 241 in samples of plutonium metal, oxide and other solid forms, when the solid is appropriately sampled and dissolved. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

Science.gov (United States)

2013-05-01

... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 1 - guideword applicability and method description.

Science.gov (United States)

Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

2008-12-15

Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design.

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)

Standard Test Method for Resin Flow of Carbon Fiber-Epoxy Prepreg

CERN Document Server

American Society for Testing and Materials. Philadelphia

1999-01-01

1.1 This test method covers the determination of the amount of resin flow that will take place from prepreg tape or sheet under given conditions of temperature and pressure. 1.2 The values stated in SI units are to be regarded as standard. The values in parentheses are for reference only. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard test method for measurement of roll wave optical distortion in heat-treated flat glass

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method is applicable to the determination of the peak-to-valley depth and peak-to-peak distances of the out-of-plane deformation referred to as roll wave which occurs in flat, heat-treated architectural glass substrates processed in a heat processing continuous or oscillating conveyance oven. 1.2 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.3 This test method does not address other flatness issues like edge kink, ream, pocket distortion, bow, or other distortions outside of roll wave as defined in this test method. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Risk prediction, safety analysis and quantitative probability methods - a caveat

International Nuclear Information System (INIS)

Critchley, O.H.

1976-01-01

Views are expressed on the use of quantitative techniques for the determination of value judgements in nuclear safety assessments, hazard evaluation, and risk prediction. Caution is urged when attempts are made to quantify value judgements in the field of nuclear safety. Criteria are given the meaningful application of reliability methods but doubts are expressed about their application to safety analysis, risk prediction and design guidances for experimental or prototype plant. Doubts are also expressed about some concomitant methods of population dose evaluation. The complexities of new designs of nuclear power plants make the problem of safety assessment more difficult but some possible approaches are suggested as alternatives to the quantitative techniques criticized. (U.K.)

Method for quantitative assessment of nuclear safety computer codes

International Nuclear Information System (INIS)

Dearien, J.A.; Davis, C.B.; Matthews, L.J.

1979-01-01

A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

Safety modelling and testing of lithium-ion batteries in electrified vehicles

Science.gov (United States)

Deng, Jie; Bae, Chulheung; Marcicki, James; Masias, Alvaro; Miller, Theodore

2018-04-01

To optimize the safety of batteries, it is important to understand their behaviours when subjected to abuse conditions. Most early efforts in battery safety modelling focused on either one battery cell or a single field of interest such as mechanical or thermal failure. These efforts may not completely reflect the failure of batteries in automotive applications, where various physical processes can take place in a large number of cells simultaneously. In this Perspective, we review modelling and testing approaches for battery safety under abuse conditions. We then propose a general framework for large-scale multi-physics modelling and experimental work to address safety issues of automotive batteries in real-world applications. In particular, we consider modelling coupled mechanical, electrical, electrochemical and thermal behaviours of batteries, and explore strategies to extend simulations to the battery module and pack level. Moreover, we evaluate safety test approaches for an entire range of automotive hardware sets from cell to pack. We also discuss challenges in building this framework and directions for its future development.

Safety Assessment for transient event occurred during the ASTS test of Hanbit Unit 2

International Nuclear Information System (INIS)

Yang, Changkeun; Kim, Yohan; Ha, Sangjun

2014-01-01

Safety Injection has been actuated during the ASTS (Automatic Seismic Trip System) test of Hanbit Unit 2 on Feb. 28, 2014. It could be bad effect on system integrity. KHNP has been performed safety assessment of system for effect of Safety Injection (SI) actuation occurred during the ASTS test of hanbit Unit 2. Stable state of nuclear power plant system has been confirmed according to Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the power plant is in a safe condition were conformed. After ASTS action, thermal elimination has been processed throughout the turbine until turbine signal occurrence because ASTS is connected to M-G set in the present hanbit Unit 2. Therefore, Safety Injection signal has been actuated by rapid reduction of Steam Generator pressure. In this paper, it is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation. Stable state of nuclear power plant system has been confirmed for Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the plant is in a safe condition were conformed. It is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation

Supplementary safety system 1/4 scale testing

Energy Technology Data Exchange (ETDEWEB)

Garrett, R.L.; Paik, I.K.

1993-09-01

During the course of updating the K-Reactor Safety Analysis Report Chapter 15 in 1990, it was identified that the current Supplementary Safety System (SSS) may not be adequate in protecting the reactor during the process water pump coastdown initiated by a loss of AC power when the safety rods are assumed to fail. A SSS modification project was initiated to add an additional ink injection pathway near the pump suction. In addition, the Department of Energy raised a question on the thermal buoyancy effects on moderator flow pattern and ink dispersion in the moderator space. The development and documentation of a two-dimensional code called MODFLOW was undertaken to describe the problem. This report discusses the results of the moderator flow and ink (Gadolinium Poison Solution - GPS) dispersion tests designed to provide qualified data for validation and benchmarking of the MODFLOW computer code with the secondary objectives being the development of concentration profiles and video footage of simulated GPS dispersion under steady-state and transient flow conditions.

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

International Nuclear Information System (INIS)

Ohi, Takao; Kawasaki, Daisuke; Chiba, Tamotsu; Takase, Toshio; Hane, Koji

2013-01-01

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner. (author)

Standard test method for determining the crevice repassivation potential of corrosion-resistant alloys using a potentiodynamic-galvanostatic-potentiostatic technique

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers a procedure for conducting anodic polarization studies to determine the crevice repassivation potential for corrosion–resistant alloys. The concept of the repassivation potential is similar to that of the protection potential given in Reference Test Method G 5. 1.2 The test method consists in applying successively potentiodynamic, galvanostatic, and potentiostatic treatments for the initial formation and afterward repassivation of crevice corrosion. 1.3 This test method is a complement to Test Method G 61. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Formal Methods and Safety Certification: Challenges in the Railways Domain

DEFF Research Database (Denmark)

Fantechi, Alessandro; Ferrari, Alessio; Gnesi, Stefania

2016-01-01

The railway signalling sector has historically been a source of success stories about the adoption of formal methods in the certification of software safety of computer-based control equipment.......The railway signalling sector has historically been a source of success stories about the adoption of formal methods in the certification of software safety of computer-based control equipment....

Standard test method for plutonium assay by plutonium (III) diode array spectrophotometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method describes the determination of total plutonium as plutonium(III) in nitrate and chloride solutions. The technique is applicable to solutions of plutonium dioxide powders and pellets (Test Methods C 697), nuclear grade mixed oxides (Test Methods C 698), plutonium metal (Test Methods C 758), and plutonium nitrate solutions (Test Methods C 759). Solid samples are dissolved using the appropriate dissolution techniques described in Practice C 1168. The use of this technique for other plutonium-bearing materials has been reported (1-5), but final determination of applicability must be made by the user. The applicable concentration range for plutonium sample solutions is 10–200 g Pu/L. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 appropria...

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

Safety test facilities. Needs and concepts. A French evaluation

International Nuclear Information System (INIS)

Tretiakoff, O.; Bailly, J.

1976-01-01

The fuel behaviour of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program (local blockage, sudden flow reduction and pump coast-down at constant power). These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The paper explains how a logical set of simple observations has led to the present state of the Cadarache in-pile experimental safety program and how it may help to find the way in a dense forest of both technical and psychological difficulties

Kartini reactor tank inspection using NDT method for safety improvement of the reactor operation

International Nuclear Information System (INIS)

Syarip; Sutondo, Tegas; Saleh, Chaerul; Nitiswati; Puradwi; Andryansah; Mudiharjo

2002-01-01

The inspection of Kartini reactor tank liner (TRK) by using Non Destructive Testing (NDT) methods to improve the reactor operation safety, have been done. The type of NDT used were: visual examination using an underwater camera and magnifier, replication survey using dental putty, hardness test using an Equotip D indentor, thickness test using ultrasonic probe, and dye penetrant test. The visual examination showed that the surface of TRK was in good condition. The hardness readings were considered to be consistent with the original condition of the tank and the slight hardness increase at the reactor core area consistent with the neutron fluence experienced -10 1 4 n/cm 2 . Results of ultrasonic thickness survey showed that in average the TRK thickness is between 5,0 mm - 6,5 mm, a low 2,1 mm thickness exists at the top of the TRK in the belt area (double layer aluminum plat, therefore do not influencing the safety ). The replica and dye penetrant test at the low thickness area and several suspected areas showed that it could be some defect from original manufacture. Therefore, it can be concluded that the TRK is still feasible for continued operation safely

Impact of proof test interval and coverage on probability of failure of safety instrumented function

International Nuclear Information System (INIS)

Jin, Jianghong; Pang, Lei; Hu, Bin; Wang, Xiaodong

2016-01-01

Highlights: • Introduction of proof test coverage makes the calculation of the probability of failure for SIF more accurate. • The probability of failure undetected by proof test is independently defined as P TIF and calculated. • P TIF is quantified using reliability block diagram and simple formula of PFD avg . • Improving proof test coverage and adopting reasonable test period can reduce the probability of failure for SIF. - Abstract: Imperfection of proof test can result in the safety function failure of safety instrumented system (SIS) at any time in its life period. IEC61508 and other references ignored or only elementarily analyzed the imperfection of proof test. In order to further study the impact of the imperfection of proof test on the probability of failure for safety instrumented function (SIF), the necessity of proof test and influence of its imperfection on system performance was first analyzed theoretically. The probability of failure for safety instrumented function resulted from the imperfection of proof test was defined as probability of test independent failures (P TIF ), and P TIF was separately calculated by introducing proof test coverage and adopting reliability block diagram, with reference to the simplified calculation formula of average probability of failure on demand (PFD avg ). Research results show that: the shorter proof test period and the higher proof test coverage indicate the smaller probability of failure for safety instrumented function. The probability of failure for safety instrumented function which is calculated by introducing proof test coverage will be more accurate.

Safety assessment and detection methods of genetically modified organisms.

Science.gov (United States)

Xu, Rong; Zheng, Zhe; Jiao, Guanglian

2014-01-01

Genetically modified organisms (GMOs), are gaining importance in agriculture as well as the production of food and feed. Along with the development of GMOs, health and food safety concerns have been raised. These concerns for these new GMOs make it necessary to set up strict system on food safety assessment of GMOs. The food safety assessment of GMOs, current development status of safety and precise transgenic technologies and GMOs detection have been discussed in this review. The recent patents about GMOs and their detection methods are also reviewed. This review can provide elementary introduction on how to assess and detect GMOs.

Zero-inflated Poisson model based likelihood ratio test for drug safety signal detection.

Science.gov (United States)

Huang, Lan; Zheng, Dan; Zalkikar, Jyoti; Tiwari, Ram

2017-02-01

In recent decades, numerous methods have been developed for data mining of large drug safety databases, such as Food and Drug Administration's (FDA's) Adverse Event Reporting System, where data matrices are formed by drugs such as columns and adverse events as rows. Often, a large number of cells in these data matrices have zero cell counts and some of them are "true zeros" indicating that the drug-adverse event pairs cannot occur, and these zero counts are distinguished from the other zero counts that are modeled zero counts and simply indicate that the drug-adverse event pairs have not occurred yet or have not been reported yet. In this paper, a zero-inflated Poisson model based likelihood ratio test method is proposed to identify drug-adverse event pairs that have disproportionately high reporting rates, which are also called signals. The maximum likelihood estimates of the model parameters of zero-inflated Poisson model based likelihood ratio test are obtained using the expectation and maximization algorithm. The zero-inflated Poisson model based likelihood ratio test is also modified to handle the stratified analyses for binary and categorical covariates (e.g. gender and age) in the data. The proposed zero-inflated Poisson model based likelihood ratio test method is shown to asymptotically control the type I error and false discovery rate, and its finite sample performance for signal detection is evaluated through a simulation study. The simulation results show that the zero-inflated Poisson model based likelihood ratio test method performs similar to Poisson model based likelihood ratio test method when the estimated percentage of true zeros in the database is small. Both the zero-inflated Poisson model based likelihood ratio test and likelihood ratio test methods are applied to six selected drugs, from the 2006 to 2011 Adverse Event Reporting System database, with varying percentages of observed zero-count cells.

New test methods for BIPV. Results from IP performance

International Nuclear Information System (INIS)

Jol, J.C.; Van Kampen, B.J.M.; De Boer, B.J.; Reil, F.; Geyer, D.

2009-11-01

Within the Performance project new test procedures for PV building products and the building performance as a whole when PV is applied in buildings have been drafted. It has resulted in a first draft of new test procedures for PV building products and proposals for tests for novel BIPV technology like thin film. The test proposed are a module breakage test for BIPV products, a fire safety test for BIPV products and a dynamic load test for BIPV products. Furthermore first proposals of how flexible PV modules could be tested in an appropriate way to ensure long time quality and safety of these new products are presented.

Results of the 1986 FFTF inherent safety tests

International Nuclear Information System (INIS)

Burke, T.M.; Campbell, L.R.; Franz, G.R.; Knecht, W.L.

1987-01-01

A series of tests was recently completed at the 400-MW (thermal) Fast Flux Test Facility (FFTF) to further demonstrate the passive safety characteristics of liquid-metal-cooled fast reactors. Earlier FFTF testing of decay heat removal by sodium natural circulation was reported in 1981. The main purpose of the 1986 test series was to demonstrate passive reactor shutdown during a loss-of-flow event when several inherent shutdown devices called gas expansion modules (GEMs) were installed in the reactor. However, these tests also provide further data on the natural circulation performance of the primary system, in particular the reactor core, and thus add to the data base available for checking the validity of available analytical tools

A simple reliability block diagram method for safety integrity verification

International Nuclear Information System (INIS)

Guo Haitao; Yang Xianhui

2007-01-01

IEC 61508 requires safety integrity verification for safety related systems to be a necessary procedure in safety life cycle. PFD avg must be calculated to verify the safety integrity level (SIL). Since IEC 61508-6 does not give detailed explanations of the definitions and PFD avg calculations for its examples, it is difficult for common reliability or safety engineers to understand when they use the standard as guidance in practice. A method using reliability block diagram is investigated in this study in order to provide a clear and feasible way of PFD avg calculation and help those who take IEC 61508-6 as their guidance. The method finds mean down times (MDTs) of both channel and voted group first and then PFD avg . The calculated results of various voted groups are compared with those in IEC61508 part 6 and Ref. [Zhang T, Long W, Sato Y. Availability of systems with self-diagnostic components-applying Markov model to IEC 61508-6. Reliab Eng System Saf 2003;80(2):133-41]. An interesting outcome can be realized from the comparison. Furthermore, although differences in MDT of voted groups exist between IEC 61508-6 and this paper, PFD avg of voted groups are comparatively close. With detailed description, the method of RBD presented can be applied to the quantitative SIL verification, showing a similarity of the method in IEC 61508-6

Methods of checking general safety criteria in UML statechart specifications

International Nuclear Information System (INIS)

Pap, Zsigmond; Majzik, Istvan; Pataricza, Andras; Szegi, Andras

2005-01-01

This paper describes methods and tools for safety analysis of UML statechart specifications. A comprehensive set of general safety criteria including completeness and consistency is applied in automated analysis. Analysis techniques are based on OCL expressions, graph transformations and reachability analysis. Two canonical intermediate representations of the statechart specification are introduced. They are suitable for straightforward implementation of checker methods and for the support of the proof of the correctness and soundness of the applied analysis. One of them also serves as a basis of the metamodel of a variant of UML statecharts proposed for the specification of safety-critical control systems. The analysis is extended to object-oriented specifications. Examples illustrate the application of the checker methods implemented by an automated tool-set

The use of human factors methods to identify and mitigate safety issues in radiation therapy

International Nuclear Information System (INIS)

Chan, Alvita J.; Islam, Mohammad K.; Rosewall, Tara; Jaffray, David A.; Easty, Anthony C.; Cafazzo, Joseph A.

2010-01-01

Background and purpose: New radiation therapy technologies can enhance the quality of treatment and reduce error. However, the treatment process has become more complex, and radiation dose is not always delivered as intended. Using human factors methods, a radiotherapy treatment delivery process was evaluated, and a redesign was undertaken to determine the effect on system safety. Material and methods: An ethnographic field study and workflow analysis was conducted to identify human factors issues of the treatment delivery process. To address specific issues, components of the user interface were redesigned through a user-centered approach. Sixteen radiation therapy students were then used to experimentally evaluate the redesigned system through a usability test to determine the effectiveness in mitigating use errors. Results: According to findings from the usability test, the redesigned system successfully reduced the error rates of two common errors (p < .04 and p < .01). It also improved the mean task completion time by 5.5% (p < .02) and achieved a higher level of user satisfaction. Conclusions: These findings demonstrated the importance and benefits of applying human factors methods in the design of radiation therapy systems. Many other opportunities still exist to improve patient safety in this area using human factors methods.

Review of studies on criticality safety evaluation and criticality experiment methods

International Nuclear Information System (INIS)

Naito, Yoshitaka; Yamamoto, Toshihiro; Misawa, Tsuyoshi; Yamane, Yuichi

2013-01-01

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS. (author)

Standard Test Methods for Insulation Integrity and Ground Path Continuity of Photovoltaic Modules

CERN Document Server

American Society for Testing and Materials. Philadelphia

2000-01-01

1.1 These test methods cover procedures for (1) testing for current leakage between the electrical circuit of a photovoltaic module and its external components while a user-specified voltage is applied and (2) for testing for possible module insulation breakdown (dielectric voltage withstand test). 1.2 A procedure is described for measuring the insulation resistance between the electrical circuit of a photovoltaic module and its external components (insulation resistance test). 1.3 A procedure is provided for verifying that electrical continuity exists between the exposed external conductive surfaces of the module, such as the frame, structural members, or edge closures, and its grounding point (ground path continuity test). 1.4 This test method does not establish pass or fail levels. The determination of acceptable or unacceptable results is beyond the scope of this test method. 1.5 There is no similar or equivalent ISO standard. This standard does not purport to address all of the safety concerns, if a...

Status of the EU test blanket systems safety studies

International Nuclear Information System (INIS)

Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

2015-01-01

Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

Status of the EU test blanket systems safety studies

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

2015-10-15

Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

Standard test method for determination of breaking strength of ceramic tiles by three-point loading

CERN Document Server

American Society for Testing and Materials. Philadelphia

2001-01-01

1.1 This test method covers the determination of breaking strength of ceramic tiles by three-point loading. 1.2 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard test methods for chemical and spectrochemical analysis of nuclear-Grade silver-indium-cadmium alloys

CERN Document Server

American Society for Testing and Materials. Philadelphia

1990-01-01

1.1 These test methods cover procedures for the chemical and spectrochemical analysis of nuclear grade silver-indium-cadmium (Ag-In-Cd) alloys to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Silver, Indium, and Cadmium by a Titration Method 7-15 Trace Impurities by Carrier-Distillation Spectro- chemical Method 16-22 1.3 The values stated in SI units are to be regarded as the standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazard and precautionary statements, see Section 5 and Practices E50. 7.1 This test method is applicable to the determination of silver, indium, and cadmium in alloys of approximately 80 % silver, 15 % indium, and 5 % cadmium used in nuclear reactor control r...

Safety measures for integrity test apparatus for IS process. Sulfuric acid decomposition section

International Nuclear Information System (INIS)

Noguchi, Hiroki; Kubo, Shinji; Iwatsuki, Jin; Onuki, Kaoru

2013-07-01

Hazardous substances such as sulfuric acid, sulfur dioxide and hydrogen iodide acid are employed in thermochemical Iodine-Sulfur (IS) process. It is necessary to take safety measure against workers and external environments to study experimentally on IS process. Presently we have been conducting to verify the soundness of main components made of engineering material in actual corrosive condition. An integrity test apparatus for the components of sulfuric acid decomposition was set up. We will use the hazardous substances such as sulfuric acid and sulfur dioxide and perform the experiment in pressurized condition in this integrity test. Safety measures for the test apparatus, operation and abnormal situation were considered prior to starting the test. This report summarized the consideration results for the safety measures on the integrity test apparatus for the components of sulfuric acid decomposition. (author)

Safety and Environment- Masterplan 2020 of DLR's Rocket Test Center Lampoldhausen

Science.gov (United States)

Haberzettl, Andreas; Dommers, Michael

2013-09-01

The German Aerospace Center DLR is the German research institute with approximately 7000 employees in 16 domestic locations. Among the research priorities of the German Aerospace Center DLR includes aerospace, energy and transport. DLR is institutionally supported by federal and state governments.Next funding sources arise in the context of third-party funds business (contract research and public contracts and subsidiaries). Main activities of the test center Lampoldshausen are testing of ARIANE's main and upper stage engines in the frame of ESA contracts.In the last years the test center of the DLR in Lampoldshausen has grown strongly, so that the number of employees is actually of about 230. The testing department is mainly responsible for rocket combustion testing according to customer requirements.Two kinds of test facilities are operated, sea level test benches and the altitude simulation test facilities.In addition to the DLR's growth also the activities of the industrial partner ASTRIUM has been elevated so that actually nearly 600 employees are present on site Lampoldshausen.The management of the site in relation to safety and security requires special measures with special respect to the presence of more people inside the testing area in order to guarantee trouble-free and safe experimental operation onsite the DLR's test plants. In order to meet with the future needs of continuing growth, the security and safety requirements have to be adopted.This report gives comprehensive outlook information about future possible scenarios of our coming tasks.Main driving force for future requests is the evolution of the rocket ARIANE. The testing of the new upper stage test facility for ARIANE 5 midlife evolution has been started. A new test position P5.2 is foreseen to perform the qualification of the new upper stage with the VINCI engine. This project will be very complex, in parallel running operation processes will require special procedures related to the overall

Safety test No. S-6, launch pad abort sequential test Phase II: solid propellant fire

International Nuclear Information System (INIS)

Snow, E.C.

1975-08-01

In preparation for the Lincoln Laboratory's LES 8/9 space mission, a series of tests was performed to evaluate the nuclear safety capability of the Multi-Hundred Watt (MHW) Radioisotope Thermoelectric Generator (RTG) to be used to supply power for the satellite. One such safety test is Test No. S-6, Launch Pad Abort Sequential Test. The objective of this test was to subject the RTG and its components to the sequential environments characteristic of a catastrophic launch pad accident to evaluate their capability to contain the 238 PuO 2 fuel. This sequence of environments was to have consisted of the blast overpressure and fragments, followed by the fireball, low velocity impact on the launch pad, and solid propellant fire. The blast overpressure and fragments were subsequently eliminated from this sequence. The procedures and results of Phase II of Test S-6, Solid Propellant Fire are presented. In this phase of the test, a simulant Fuel Sphere Assembly (FSA) and a mockup of a damaged Heat Source Assembly (HSA) were subjected to single proximity solid propellant fires of approximately 10-min duration. Steel was introduced into both tests to simulate the effects of launch pad debris and the solid rocket motor (SRM) casing that might be present in the fire zone. (TFD)

Tests for validation of fast neutron reactors safety

International Nuclear Information System (INIS)

Nagata, T.; Yamashita, H.

2001-01-01

Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

Survey of systems safety analysis methods and their application to nuclear waste management systems

International Nuclear Information System (INIS)

Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

1981-11-01

This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study

Survey of systems safety analysis methods and their application to nuclear waste management systems

Energy Technology Data Exchange (ETDEWEB)

Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

1981-11-01

This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

Application of reliability analysis methods to the comparison of two safety circuits

International Nuclear Information System (INIS)

Signoret, J.-P.

1975-01-01

Two circuits of different design, intended for assuming the ''Low Pressure Safety Injection'' function in PWR reactors are analyzed using reliability methods. The reliability analysis of these circuits allows the failure trees to be established and the failure probability derived. The dependence of these results on test use and maintenance is emphasized as well as critical paths. The great number of results obtained may allow a well-informed choice taking account of the reliability wanted for the type of circuits [fr

A Test Suite for Safety-Critical Java using JML

DEFF Research Database (Denmark)

Ravn, Anders Peter; Søndergaard, Hans

2013-01-01

Development techniques are presented for a test suite for the draft specification of the Java profile for Safety-Critical Systems. Distinguishing features are: specification of conformance constraints in the Java Modeling Language, encoding of infrastructure concepts without implementation bias......, and corresponding specifications of implicitly stated behavioral and real-time properties. The test programs are auto-generated from the specification, while concrete values for test parameters are selected manually. The suite is open source and publicly accessible....

Validation of calculational methods for nuclear criticality safety - approved 1975

International Nuclear Information System (INIS)

Anon.

1977-01-01

The American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, N16.1-1975, states in 4.2.5: In the absence of directly applicable experimental measurements, the limits may be derived from calculations made by a method shown to be valid by comparison with experimental data, provided sufficient allowances are made for uncertainties in the data and in the calculations. There are many methods of calculation which vary widely in basis and form. Each has its place in the broad spectrum of problems encountered in the nuclear criticality safety field; however, the general procedure to be followed in establishing validity is common to all. The standard states the requirements for establishing the validity and area(s) of applicability of any calculational method used in assessing nuclear criticality safety

Standard Test Method for Thermal Oxidative Resistance of Carbon Fibers

CERN Document Server

American Society for Testing and Materials. Philadelphia

1982-01-01

1.1 This test method covers the apparatus and procedure for the determination of the weight loss of carbon fibers, exposed to ambient hot air, as a means of characterizing their oxidative resistance. 1.2 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to inch-pound units which are provided for information only and are not considered standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazard information, see Section 8.

Tiered guidance for risk-informed environmental health and safety testing of nanotechnologies

International Nuclear Information System (INIS)

Collier, Zachary A.; Kennedy, Alan J.; Poda, Aimee R.; Cuddy, Michael F.; Moser, Robert D.; MacCuspie, Robert I.; Harmon, Ashley; Plourde, Kenton; Haines, Christopher D.; Steevens, Jeffery A.

2015-01-01

Provided the rapid emergence of novel technologies containing engineered nanomaterials, there is a need to better understand the potential environmental, health, and safety effects of nanotechnologies before wide-scale deployment. However, the unique properties of nanomaterials and uncertainty regarding applicable test methods have led to a lack of consensus regarding the collection and evaluation of data related to hazard and exposure potentials. Often, overly conservative approaches to characterization and data collection result in prolonged, unfocused, or irrelevant testing, which increases costs and delays deployment. In this paper, we provide a novel testing guidance framework for determining whether a nanotechnology has the potential to release material with nano-specific parameters that pose a risk to humans or the environment. The framework considers methods to categorize nanotechnologies by their structure and within their relevant-use scenarios to inform testing in a time- and resource-limited reality. Based on the precedent of dredged sediment testing, a five-tiered approach is proposed in which opportunities are presented to conclude testing once sufficient risk-related information has been collected, or that the technology in question does not require nano-specific scrutiny. A series of screening stages are suggested, covering relevant aspects including size, surface area, distribution, unique behaviors, and release potential. The tiered, adaptive guidance approach allows users to concentrate on collecting the most relevant data, thus accelerating technology deployment while minimizing risk

Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. 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)

Standard Test Method for Measuring Heat Flux Using a Water-Cooled Calorimeter

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 This test method covers the measurement of a steady heat flux to a given water-cooled surface by means of a system energy balance. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Potential safety enhancements to nuclear plant control: proof testing at EBR-II

International Nuclear Information System (INIS)

Lindsay, R.W.; Chisholm, G.H.

1984-01-01

Future changes in nuclear plant control and protective systems will reflect an evolutionary improvement through increased use of computers coupled with a better integration of man and machine. Before improvements can be accepted into the licensed commercial plant environment, significant testing must be accomplished to answer safety questions and to prove the worth of new ideas. The Experimental Breeder Reactor-II (EBR-II) is being used as a test-bed for both in-house development and testing for others in a DOE sponsored Man-Machine Integration program. The ultimate result of the development and testing would be a control system for which safety credit could be taken in the licensing process

The discussion on the qualitative and quantitative evaluation methods for safety culture

International Nuclear Information System (INIS)

Gao Kefu

2005-01-01

The fundamental methods for safely culture evaluation are described. Combining with the practice of the quantitative evaluation of safety culture in Daya Bay NPP, the quantitative evaluation method for safety culture are discussed. (author)

NRC confirmatory safety system testing in support of AP600 design review

International Nuclear Information System (INIS)

Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.

1994-01-01

Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation

77 FR 10666 - Pipeline Safety: Post Accident Drug and Alcohol Testing

Science.gov (United States)

2012-02-23

... 199 [Docket No. PHMSA-2011-0335] Pipeline Safety: Post Accident Drug and Alcohol Testing AGENCY... operators of Liquefied Natural Gas (LNG) facilities to conduct post- accident drug and alcohol tests of..., operators must drug and alcohol test each covered employee whose performance either contributed to the...

A tiered approach to the use of alternatives to animal testing for the safety assessment of cosmetics: skin irritation.

Science.gov (United States)

Macfarlane, Martin; Jones, Penny; Goebel, Carsten; Dufour, Eric; Rowland, Joanna; Araki, Daisuke; Costabel-Farkas, Margit; Hewitt, Nicola J; Hibatallah, Jalila; Kirst, Annette; McNamee, Pauline; Schellauf, Florian; Scheel, Julia

2009-07-01

Evaluation of the skin irritancy and corrosivity potential of an ingredient is a necessity in the safety assessment of cosmetic ingredients. To date, there are two formally validated alternatives to the rabbit Draize test for skin corrosivity in place, namely the rat skin transcutaneous electrical resistance (TER) assay and the Human Skin Model Test using EpiSkin, EpiDerm and SkinEthic reconstructed human epidermal equivalents. For skin irritation, EpiSkin, EpiDerm and SkinEthic are validated as stand-alone test replacements for the rabbit Draize test. Data from these tests are rarely considered in isolation and are evaluated in combination with other factors to establish the overall irritating or corrosive potential of an ingredient. In light of the deadlines established in the Cosmetics Directive for cessation of animal testing for cosmetic ingredients, a COLIPA scientific meeting was held in Brussels on 30th January, 2008 to review the use of alternative approaches and to set up a decision tree approach for their integration into tiered testing strategies for hazard and safety assessment of cosmetic ingredients and their use in products. In conclusion, the safety assessments for skin irritation/corrosion of new chemicals for use in cosmetics can be confidently accomplished using exclusively alternative methods.

Results and implications of the EBR-II inherent safety demonstration tests

International Nuclear Information System (INIS)

Planchon, H.P.; Golden, G.H.; Sackett, J.I.; Mohr, D.; Chang, L.K.; Feldman, E.E.; Betten, P.R.

1987-01-01

On April 3, 1986 two milestone tests were conducted in Experimental Breeder Reactor-2 (EBR-II). The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power and in both tests the reactor was shut down by natural processes, principally thermal expansion, without automatic scram, operator intervention or the help of special in-core devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. In a general sense, therefore, the tests plus supporting analysis demonstrated the feasibility of inherent passive shutdown for undercooling accidents in metal-fueled LMRs. The results provide a technical basis for future experiments in EBR-II to demonstrate inherent safety for overpower accidents and provide data for validation of computer codes used for design and safety analysis of inherently safe reactor plants

Field Test of the World Health Organization Multi-Professional Patient Safety Curriculum Guide.

Science.gov (United States)

Farley, Donna; Zheng, Hao; Rousi, Eirini; Leotsakos, Agnès

2015-01-01

Although the importance of training in patient safety has been acknowledged for over a decade, it remains under-utilized and under-valued in most countries. WHO developed the Multi-professional Patient Safety Curriculum Guide to provide schools with the requirements and tools for incorporating patient safety in education. It was field tested with 12 participating schools across the six WHO regions, to assess its effectiveness for teaching patient safety to undergraduate and graduate students in a global variety of settings. The evaluation used a combined prospective/retrospective design to generate formative information on the experiences of working with the Guide and summative information on the impacts of the Guide. Using stakeholder interviews and student surveys, data were gathered from each participating school at three times: the start of the field test (baseline), soon after each school started teaching, and soon after each school finished teaching. Stakeholders interviewed were strongly positive about the Guide, noting that it emphasized universally important patient safety topics, was culturally appropriate for their countries, and gave credibility and created a focus on patient safety at their schools. Student perceptions and attitudes regarding patient safety improved substantially during the field test, and their knowledge of the topics they were taught doubled, from 10.7% to 20.8% of correct answers on the student survey. This evaluation documented the effectiveness of the Curriculum Guide, for both ease of use by schools and its impacts on improving the patient safety knowledge of healthcare students. WHO should be well positioned to refine the contents of the Guide and move forward in encouraging broader use of the Guide globally for teaching patient safety.

Method of operator safety assessment for underground mobile mining equipment

Science.gov (United States)

Działak, Paulina; Karliński, Jacek; Rusiński, Eugeniusz

2018-01-01

The paper presents a method of assessing the safety of operators of mobile mining equipment (MME), which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM). As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

Method of operator safety assessment for underground mobile mining equipment

Directory of Open Access Journals (Sweden)

Działak Paulina

2018-01-01

Full Text Available The paper presents a method of assessing the safety of operators of mobile mining equipment (MME, which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM. As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

International Nuclear Information System (INIS)

Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

1992-07-01

This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

Seismic safety margin assessment program (Annual safety research report, JFY 2010)

International Nuclear Information System (INIS)

Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

2011-01-01

Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

[Agricultural biotechnology safety assessment].

Science.gov (United States)

McClain, Scott; Jones, Wendelyn; He, Xiaoyun; Ladics, Gregory; Bartholomaeus, Andrew; Raybould, Alan; Lutter, Petra; Xu, Haibin; Wang, Xue

2015-01-01

Genetically modified (GM) crops were first introduced to farmers in 1995 with the intent to provide better crop yield and meet the increasing demand for food and feed. GM crops have evolved to include a thorough safety evaluation for their use in human food and animal feed. Safety considerations begin at the level of DNA whereby the inserted GM DNA is evaluated for its content, position and stability once placed into the crop genome. The safety of the proteins coded by the inserted DNA and potential effects on the crop are considered, and the purpose is to ensure that the transgenic novel proteins are safe from a toxicity, allergy, and environmental perspective. In addition, the grain that provides the processed food or animal feed is also tested to evaluate its nutritional content and identify unintended effects to the plant composition when warranted. To provide a platform for the safety assessment, the GM crop is compared to non-GM comparators in what is typically referred to as composition equivalence testing. New technologies, such as mass spectrometry and well-designed antibody-based methods, allow better analytical measurements of crop composition, including endogenous allergens. Many of the analytical methods and their intended uses are based on regulatory guidance documents, some of which are outlined in globally recognized documents such as Codex Alimentarius. In certain cases, animal models are recommended by some regulatory agencies in specific countries, but there is typically no hypothesis or justification of their use in testing the safety of GM crops. The quality and standardization of testing methods can be supported, in some cases, by employing good laboratory practices (GLP) and is recognized in China as important to ensure quality data. Although the number of recommended, in some cases, required methods for safety testing are increasing in some regulatory agencies, it should be noted that GM crops registered to date have been shown to be

Safety Evaluation of Radioactive Material Transport Package under Stacking Test Condition

International Nuclear Information System (INIS)

Lee, Ju Chan; Seo, Ki Seog; Yoo, Seong Yeon

2012-01-01

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

76 FR 34145 - Safety Zone, Barrier Testing Operations, Chicago Sanitary and Ship Canal, Romeoville, IL

Science.gov (United States)

2011-06-13

...-AA00 Safety Zone, Barrier Testing Operations, Chicago Sanitary and Ship Canal, Romeoville, IL AGENCY.... Construction on Barrier IIB has been completed. Operational and safety testing was conducted in February 2011... dispersal barrier IIA and IIB. This safety zone will be enforced daily from 7 a.m. to 11 a.m. and from 1 p.m...

Method and practice on safety software verification and validation for digital reactor protection system

International Nuclear Information System (INIS)

Li Duo; Zhang Liangju; Feng Junting

2010-01-01

The key issue arising from digitalization of reactor protection system for Nuclear Power Plant (NPP) is in essence, how to carry out Verification and Validation (V and V), to demonstrate and confirm the software is reliable enough to perform reactor safety functions. Among others the most important activity of software V and V process is unit testing. This paper discusses the basic concepts on safety software V and V and the appropriate technique for software unit testing, focusing on such aspects as how to ensure test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed herein was successfully used in the work of unit testing on safety software of a digital reactor protection system. (author)

Standard Test Methods for Properties of Continuous Filament Carbon and Graphite Fiber Tows

CERN Document Server

American Society for Testing and Materials. Philadelphia

1999-01-01

1.1 These test methods cover the preparation and tensile testing of resin-impregnated and consolidated test specimens made from continuous filament carbon and graphite yarns, rovings, and tows to determine their tensile properties. 1.2 These test methods also cover the determination of the density and mass per unit length of the yarn, roving, or tow to provide supplementary data for tensile property calculation. 1.3 These test methods include a procedure for sizing removal to provide the preferred desized fiber samples for density measurement. This procedure may also be used to determine the weight percent sizing. 1.4 These test methods include a procedure for determining the weight percent moisture adsorption of carbon or graphite fiber. 1.5 The values stated in SI units are to be regarded as the standard. The values in parentheses are for information only. 1.6 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 t...

Standard Test Method for Stress-Corrosion of Titanium Alloys by Aircraft Engine Cleaning Materials

CERN Document Server

American Society for Testing and Materials. Philadelphia

2006-01-01

1.1 This test method establishes a test procedure for determining the propensity of aircraft turbine engine cleaning and maintenance materials for causing stress corrosion cracking of titanium alloy parts. 1.2 The evaluation is conducted on representative titanium alloys by determining the effect of contact with cleaning and maintenance materials on tendency of prestressed titanium alloys to crack when subsequently heated to elevated temperatures. 1.3 Test conditions are based upon manufacturer's maximum recommended operating solution concentration. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific precautionary statements, see and .

Analysis Method of Common Cause Failure on Non-safety Digital Control System

Energy Technology Data Exchange (ETDEWEB)

Kim, Yun Goo; Oh, Eun Gse [KHNP, Daejeon (Korea, Republic of)

2014-08-15

The effects of common cause failure on safety digital instrumentation and control system had been considered in defense in depth analysis with safety analysis method. However, the effects of common cause failure on non-safety digital instrumentation and control system also should be evaluated. The common cause failure can be included in credible failure on the non-safety system. In the I and C architecture of nuclear power plant, many design feature has been applied for the functional integrity of control system. One of that is segmentation. Segmentation defenses the propagation of faults in the I and C architecture. Some of effects from common cause failure also can be limited by segmentation. Therefore, in this paper there are two type of failure mode, one is failures in one control group which is segmented, and the other is failures in multiple control group because that the segmentation cannot defense all effects from common cause failure. For each type, the worst failure scenario is needed to be determined, so the analysis method has been proposed in this paper. The evaluation can be qualitative when there is sufficient justification that the effects are bounded in previous safety analysis. When it is not bounded in previous safety analysis, additional analysis should be done with conservative assumptions method of previous safety analysis or best estimation method with realistic assumptions.

On Some Methods in Safety Evaluation in Geotechnics

Science.gov (United States)

Puła, Wojciech; Zaskórski, Łukasz

2015-06-01

The paper demonstrates how the reliability methods can be utilised in order to evaluate safety in geotechnics. Special attention is paid to the so-called reliability based design that can play a useful and complementary role to Eurocode 7. In the first part, a brief review of first- and second-order reliability methods is given. Next, two examples of reliability-based design are demonstrated. The first one is focussed on bearing capacity calculation and is dedicated to comparison with EC7 requirements. The second one analyses a rigid pile subjected to lateral load and is oriented towards working stress design method. In the second part, applications of random field to safety evaluations in geotechnics are addressed. After a short review of the theory a Random Finite Element algorithm to reliability based design of shallow strip foundation is given. Finally, two illustrative examples for cohesive and cohesionless soils are demonstrated.

[Experience feedback committee: a method for patient safety improvement].

Science.gov (United States)

François, P; Sellier, E; Imburchia, F; Mallaret, M-R

2013-04-01

An experience feedback committee (CREX, Comité de Retour d'EXpérience) is a method which contributes to the management of safety of care in a medical unit. Originally used for security systems of civil aviation, the method has been adapted to health care facilities and successfully implemented in radiotherapy units and in other specialties. We performed a brief review of the literature for studies reporting data on CREX established in hospitals. The review was performed using the main bibliographic databases and Google search results. The CREX is designed to analyse incidents reported by professionals. The method includes monthly meetings of a multi-professional committee that reviews the reported incidents, chooses a priority incident and designates a "pilot" responsible for investigating the incident. The investigation of the incident involves a systemic analysis method and a written synthesis presented at the next meeting of the committee. The committee agrees on actions for improvement that are suggested by the analysis and follows their implementation. Systems for the management of health care, including reporting systems, are organized into three levels: the medical unit, the hospital and the country as a triple loop learning process. The CREX is located in the base level, short loop of risk management and allows direct involvement of care professionals in patient safety. Safety of care has become a priority of health systems. In this context, the CREX can be a useful vehicle for the implementation of a safety culture in medical units. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

Efficient improvement of nuclear power plant safety by reorganization of risk-informed safety importance evaluation methods for piping welded portions

Energy Technology Data Exchange (ETDEWEB)

Irie, Takashi; Hanafusa, Hidemitsu; Suyama, Takeshi [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Morota, Hidetsugu; Kojima, Sigeo; Mizuno, Yoshinobu [Computer Software Development Co., Ltd., Tokyo (Japan)

2002-09-01

In this work, risk information was used to evaluate the safety importance of piping welded portions which were important for plant operation and maintenance of nuclear power plants. There are two types of risk-informed safety importance evaluation methods, namely the ASME method and the EPRI method. Since both methods have advantages and disadvantages, elements of each method were combined and reorganized. Considerations included whether the degradation mechanisms would be objectively evaluated and whether plant safety would be efficiently improved. The most objective and efficient method was as follows. Piping failure potential is quantitatively and objectively evaluated for failure with probabilistic fracture mechanics (PFM) and for other degradation mechanisms with empirical failure rates, and conditional core damage probability (CCDP) is calculated with PSA. This method reduces the inspected segment numbers to 1/4 of the deterministic method and increases the ratio of risk, which is covered by the inspected segments, to total risk from 80% of the deterministic method to 95%. Piping inspection numbers decreased for safety injection systems that were required the inspections by the deterministic method. Piping inspections were required for part of main feed water and main steam systems that were not required the inspections by the deterministic method. (author)

Cluster monte carlo method for nuclear criticality safety calculation

International Nuclear Information System (INIS)

Pei Lucheng

1984-01-01

One of the most important applications of the Monte Carlo method is the calculation of the nuclear criticality safety. The fair source game problem was presented at almost the same time as the Monte Carlo method was applied to calculating the nuclear criticality safety. The source iteration cost may be reduced as much as possible or no need for any source iteration. This kind of problems all belongs to the fair source game prolems, among which, the optimal source game is without any source iteration. Although the single neutron Monte Carlo method solved the problem without the source iteration, there is still quite an apparent shortcoming in it, that is, it solves the problem without the source iteration only in the asymptotic sense. In this work, a new Monte Carlo method called the cluster Monte Carlo method is given to solve the problem further

Nuclear safety

International Nuclear Information System (INIS)

Tarride, Bruno

2015-10-01

The author proposes an overview of methods and concepts used in the nuclear industry, at the design level as well as at the exploitation level, to ensure an acceptable safety level, notably in the case of nuclear reactors. He first addresses the general objectives of nuclear safety and the notion of acceptable risk: definition and organisation of nuclear safety (relationships between safety authorities and operators), notion of acceptable risk, deterministic safety approach and main safety principles (safety functions and confinement barriers, concept of defence in depth). Then, the author addresses the safety approach at the design level: studies of operational situations, studies of internal and external aggressions, safety report, design principles for important-for-safety systems (failure criterion, redundancy, failure prevention, safety classification). The next part addresses safety during exploitation and general exploitation rules: definition of the operation domain and of its limits, periodic controls and tests, management in case of incidents, accidents or aggressions

Centrifuge model test of rock slope failure caused by seismic excitation. Applicability to the stability evaluation method of safety factors against sliding

International Nuclear Information System (INIS)

Ishimaru, Makoto; Kawai, Tadashi

2010-01-01

The purposes of this study are to analyze dynamic failure characteristics of slopes in discontinuous rock mass with brittle fracture by centrifuge model tests and to study applicability to the equivalent linear analysis against dynamic sliding failure of rock slopes. We conducted centrifuge model test using a dip slope model with discontinuities imitated by Teflon sheets. The centrifugal acceleration was 30G, and the acceleration amplitudes of input sin waves were increased gradually at every step. The test results were compared with safety factors of the sliding surface based on the equivalent linear analysis. The following results were obtained: (1) The slope model collapsed when it was excited by the sine wave of 350gal, which was converted to real field scale. (2) Artificial discontinuities considerably affected the collapse, and the type of collapse was plane failure. (3) From response displacement records measured at the slope model, the failure around toe of the slope model probably caused the collapse. (4) The evaluation of safety factors against sliding based on the equivalent linear analysis were conservative compared with the experimental results. (author)

The development and standardization of testing methods for genetically modified organisms and their derived products.

Science.gov (United States)

Zhang, Dabing; Guo, Jinchao

2011-07-01

As the worldwide commercialization of genetically modified organisms (GMOs) increases and consumers concern the safety of GMOs, many countries and regions are issuing labeling regulations on GMOs and their products. Analytical methods and their standardization for GM ingredients in foods and feed are essential for the implementation of labeling regulations. To date, the GMO testing methods are mainly based on the inserted DNA sequences and newly produced proteins in GMOs. This paper presents an overview of GMO testing methods as well as their standardization. © 2011 Institute of Botany, Chinese Academy of Sciences.

Verification of Overall Safety Factors In Deterministic Design Of Model Tested Breakwaters

DEFF Research Database (Denmark)

Burcharth, H. F.

2001-01-01

The paper deals with concepts of safety implementation in design. An overall safety factor concept is evaluated on the basis of a reliability analysis of a model tested rubble mound breakwater with monolithic super structure. Also discussed are design load identification and failure mode limit...

Development of partitioning method : cold experiment with partitioning test facility in NUCEF (I)

International Nuclear Information System (INIS)

Yamaguchi, Isoo; Morita, Yasuji; Kondo, Yasuo

1996-03-01

A test facility in which about 1.85 x 10 14 Bq of high-level liquid waste can be treated has been completed in 1994 at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) for research and development of Partitioning Method. The outline of the partitioning test facility and support equipments for it which were design terms, constructions, arrangements, functions and inspections were given in JAERI-Tech 94-030. The present report describes the results of the water transfer test and partitioning tests, which are methods of precipitation by denitration, oxalate precipitation, solvent extraction, and adsorption with inorganic ion exchanger, using nitric acid to master operation method of the test facility. As often as issues related to equipments occurred during the tests, they were improved. As to issues related to processes such as being stopped up of columns, their measures of solution were found by testing in laboratories. They were reflected in operation of the Partitioning Test Facility. Their particulars and improving points were described in this report. (author)

Advances in methods and applications of reliability and safety analysis

International Nuclear Information System (INIS)

Fieandt, J.; Hossi, H.; Laakso, K.; Lyytikaeinen, A.; Niemelae, I.; Pulkkinen, U.; Pulli, T.

1986-01-01

The know-how of the reliability and safety design and analysis techniques of Vtt has been established over several years in analyzing the reliability in the Finnish nuclear power plants Loviisa and Olkiluoto. This experience has been later on applied and developed to be used in the process industry, conventional power industry, automation and electronics. VTT develops and transfers methods and tools for reliability and safety analysis to the private and public sectors. The technology transfer takes place in joint development projects with potential users. Several computer-aided methods, such as RELVEC for reliability modelling and analysis, have been developed. The tool developed are today used by major Finnish companies in the fields of automation, nuclear power, shipbuilding and electronics. Development of computer-aided and other methods needed in analysis of operating experience, reliability or safety is further going on in a number of research and development projects

Standard test method for nondestructive assay of special nuclear material holdup using Gamma-Ray spectroscopic methods

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U, or 239Pu remaining as holdup in nuclear facilities. Holdup occurs in all facilities where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors. 1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources (1, 2, 3, 4). 1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements; as well as the laws of physics. The measurement process includes defining measurement uncertainties and is sensitive to the form and distribution of the material...

Environmental assessment of general-purpose heat source safety verification testing

International Nuclear Information System (INIS)

1995-02-01

This Environmental Assessment (EA) was prepared to identify and evaluate potential environmental, safety, and health impacts associated with the Proposed Action to test General-Purpose Heat Source (GPHS) Radioisotope Thermoelectric Generator (RTG) assemblies at the Sandia National Laboratories (SNL) 10,000-Foot Sled Track Facility, Albuquerque, New Mexico. RTGs are used to provide a reliable source of electrical power on board some spacecraft when solar power is inadequate during long duration space missions. These units are designed to convert heat from the natural decay of radioisotope fuel into electrical power. Impact test data are required to support DOE's mission to provide radioisotope power systems to NASA and other user agencies. The proposed tests will expand the available safety database regarding RTG performance under postulated accident conditions. Direct observations and measurements of GPHS/RTG performance upon impact with hard, unyielding surfaces are required to verify model predictions and to ensure the continual evolution of the RTG designs that perform safely under varied accident environments. The Proposed Action is to conduct impact testing of RTG sections containing GPHS modules with simulated fuel. End-On and Side-On impact test series are planned

Safety instrumented systems in the oil and gas industry : Concepts and methods for safety and reliability assessments in design and operation

Energy Technology Data Exchange (ETDEWEB)

Lundteigen, Mary Ann

2009-07-01

time, they are requested to use the safety life cycle model in IEC 61508 as basis for their product development. This thesis links the safety life cycle model in IEC 61508 to a more general product development model, where IEC 61508 requirements are discussed in light of other RAMS requirements. SIS manufacturers who develop products for more than one industry sector must often adhere to IEC 61508 as well as sector specific standards. Some of the sector specific standards build directly on IEC 61508, while others have been developed prior to IEC 61508 and may use different concepts and approaches. This work describes an approach for qualification of a software development platform in light of these challenges. SIS designers have to balance the SIS reliability with the practicality of performing functional tests. Functional tests are important means to reveal SIS failures, but the tests often require process shutdowns. Partial stroke testing is a valve test that does not require full valve closure, but the test efficiency is highly influenced by the test coverage. This thesis suggests a way to determine the test coverage, taking into account application specific considerations and generic data for valve failures. The oil and gas industry has aimed at keeping the SIS as simple as possible, and here, reliability block diagrams have been well suited for reliability analysis. However, technology development challenges the principle of simplicity, and then fault tree analysis may be better suited to achieve complete reliability models and for involving design engineers in the model construction and verification. Many software tools for fault tree analysis make non conservative estimates for the reliability of periodically tested SISs, which may not be acceptable when the estimates are used as basis for selecting hardware architecture. This thesis proposes a conservative calculation approach for fault tree analysis that builds on calculation methods familiar to the oil and

Environmental tests of a digital safety channel: An investigation of stress-related vulnerabilities of computer-based safety system

International Nuclear Information System (INIS)

Korsah, K.; Wilson, T.L.; Wood, R.; Tanaka, T.

1997-01-01

This article presents the results of environmental stress tests performed on an experimental digital safety channel (EDSC) assembled at the Oak Ridge National Laboratory as part of the Qualification of Advanced Instrumentation and Controls Systems Research program, which was sponsored by the US Nuclear Regulatory Commission. The program is expected to provide recommendations for environmental qualification of digital safety systems. The purpose of the study was to investigate potential vulnerabilities of distributed computer systems used in safety applications when subjected to environmental stressors. The EDSC assembled for the tests employs technologies and digital subsystems representative of those proposed for use in advanced light-water reactors or as retrofits in existing plants. Subsystems include computers, electrical and optical serial communication links, fiber-optic network links, analog-to-digital and digital-to-analog converters, and multiplexers. The EDSC was subjected to selected stressors that are a potential risk to digital equipment in a mild environment. The selected stressors were electromagnetic and radiofrequency interferences (EMI-RFI), temperature, humidity, and smoke exposure. The stressors were applied at levels of intensity considerably higher than the safety channel is likely to experience in a normal nuclear power plant environment. Ranges of stress were selected at a sufficiently high level to induce errors so that failure modes that are characteristic of the technologies employed could be identified. On the basis of the incidence of functional errors observed during testing, EMI-RFI, smoke exposure, and high temperature coupled with high relative humidity, in that order, were found to have the greatest impact of the stressors investigated. The most prevalent stressor-induced upsets, as well as the most severe, were found to occur during the EMI-RFI tests

Standards for radiation protection instrumentation: design of safety standards and testing procedures

International Nuclear Information System (INIS)

Meissner, Frank

2008-01-01

This paper describes by means of examples the role of safety standards for radiation protection and the testing and qualification procedures. The development and qualification of radiation protection instrumentation is a significant part of the work of TUV NORD SysTec, an independent expert organisation in Germany. The German Nuclear Safety Standards Commission (KTA) establishes regulations in the field of nuclear safety. The examples presented may be of importance for governments and nuclear safety authorities, for nuclear operators and for manufacturers worldwide. They demonstrate the advantage of standards in the design of radiation protection instrumentation for new power plants, in the upgrade of existing instrumentation to nuclear safety standards or in the application of safety standards to newly developed equipment. Furthermore, they show how authorities may proceed when safety standards for radiation protection instrumentation are not yet established or require actualization. (author)

Standard test method for creep-fatigue testing

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This test method covers the determination of mechanical properties pertaining to creep-fatigue deformation or crack formation in nominally homogeneous materials, or both by the use of test specimens subjected to uniaxial forces under isothermal conditions. It concerns fatigue testing at strain rates or with cycles involving sufficiently long hold times to be responsible for the cyclic deformation response and cycles to crack formation to be affected by creep (and oxidation). It is intended as a test method for fatigue testing performed in support of such activities as materials research and development, mechanical design, process and quality control, product performance, and failure analysis. The cyclic conditions responsible for creep-fatigue deformation and cracking vary with material and with temperature for a given material. 1.2 The use of this test method is limited to specimens and does not cover testing of full-scale components, structures, or consumer products. 1.3 This test method is primarily ...

Demonstration tests for low level radioactive waste packaging safety

International Nuclear Information System (INIS)

Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

1993-01-01

The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

On Some Methods in Safety Evaluation in Geotechnics

Directory of Open Access Journals (Sweden)

Puła Wojciech

2015-06-01

Full Text Available The paper demonstrates how the reliability methods can be utilised in order to evaluate safety in geotechnics. Special attention is paid to the so-called reliability based design that can play a useful and complementary role to Eurocode 7. In the first part, a brief review of first- and second-order reliability methods is given. Next, two examples of reliability-based design are demonstrated. The first one is focussed on bearing capacity calculation and is dedicated to comparison with EC7 requirements. The second one analyses a rigid pile subjected to lateral load and is oriented towards working stress design method. In the second part, applications of random field to safety evaluations in geotechnics are addressed. After a short review of the theory a Random Finite Element algorithm to reliability based design of shallow strip foundation is given. Finally, two illustrative examples for cohesive and cohesionless soils are demonstrated.

Safety equipment and methods for evaluating its effectiveness

Energy Technology Data Exchange (ETDEWEB)

Evdokimov, F I; Nadtoka, T B [DPI (Ukraine)

1993-05-01

Analyzes relations between technologies (especially for roof support) used in black coal mining and work safety in mines. The share of manual work and accident rate are compared for mining by narrow and wide web shearer loaders and by coal plows with powered and individual support. Protection from occupational injury is discussed at three levels: safety engineering, work organization and the human factor. A method of evaluating the social and economic effectiveness of protection from occupational injury developed at the DPI institute is presented. The method uses the knowledge of probability distribution of failure situations, failures and protective means to determine the probabilistic characteristics of the functioning of protection systems and to calculate, for a given period, the occurrence probability and mean number of accidents. Each state of the system is characterized by determined social and/or economic results. The method was used in designing equipment intended for protective power cut-off in electric mine networks.

US Department of Energy DOE Nevada Operations Office, Nevada Test Site: Underground safety and health standards

Energy Technology Data Exchange (ETDEWEB)

1993-05-01

The Nevada Test Site Underground Safety and Health Standards Working Group was formed at the direction of John D. Stewart, Director, Nevada Test Site Office in April, 1990. The objective of the Working Group was to compile a safety and health standard from the California Tunnel Safety Orders and OSHA for the underground operations at the NTS, (excluding Yucca Mountain). These standards are called the NTS U/G Safety and Health Standards. The Working Group submits these standards as a RECOMMENDATION to the Director, NTSO. Although the Working Group considers these standards to be the most integrated and comprehensive standards that could be developed for NTS Underground Operations, the intent is not to supersede or replace any relevant DOE orders. Rather the intent is to collate the multiple safety and health references contained in DOE Order 5480.4 that have applicability to NTS Underground Operations into a single safety and heath standard to be used in the underground operations at the NTS. Each portion of the standard was included only after careful consideration by the Working Group and is judged to be both effective and appropriate. The specific methods and rationale used by the Working Group are outlined as follows: The letter from DOE/HQ, dated September 28, 1990 cited OSHA and the CTSO as the safety and health codes applicable to underground operations at the NTS. These mandated codes were each originally developed to be comprehensive, i.e., all underground operations of a particular type (e.g., tunnels in the case of the CTSO) were intended to be adequately regulated by the appropriate code. However, this is not true; the Working Group found extensive and confusing overlap in the codes in numerous areas. Other subjects and activities were addressed by the various codes in cursory fashion or not at all.

US Department of Energy DOE Nevada Operations Office, Nevada Test Site: Underground safety and health standards

International Nuclear Information System (INIS)

1993-05-01

The Nevada Test Site Underground Safety and Health Standards Working Group was formed at the direction of John D. Stewart, Director, Nevada Test Site Office in April, 1990. The objective of the Working Group was to compile a safety and health standard from the California Tunnel Safety Orders and OSHA for the underground operations at the NTS, (excluding Yucca Mountain). These standards are called the NTS U/G Safety and Health Standards. The Working Group submits these standards as a RECOMMENDATION to the Director, NTSO. Although the Working Group considers these standards to be the most integrated and comprehensive standards that could be developed for NTS Underground Operations, the intent is not to supersede or replace any relevant DOE orders. Rather the intent is to collate the multiple safety and health references contained in DOE Order 5480.4 that have applicability to NTS Underground Operations into a single safety and heath standard to be used in the underground operations at the NTS. Each portion of the standard was included only after careful consideration by the Working Group and is judged to be both effective and appropriate. The specific methods and rationale used by the Working Group are outlined as follows: The letter from DOE/HQ, dated September 28, 1990 cited OSHA and the CTSO as the safety and health codes applicable to underground operations at the NTS. These mandated codes were each originally developed to be comprehensive, i.e., all underground operations of a particular type (e.g., tunnels in the case of the CTSO) were intended to be adequately regulated by the appropriate code. However, this is not true; the Working Group found extensive and confusing overlap in the codes in numerous areas. Other subjects and activities were addressed by the various codes in cursory fashion or not at all

Using partial safety factors in wind turbine design and testing

Energy Technology Data Exchange (ETDEWEB)

Musial, W.D. [National Renewable Energy Lab., Golden, CO (United States)

1997-12-31

This paper describes the relationship between wind turbine design and testing in terms of the certification process. An overview of the current status of international certification is given along with a description of limit-state design basics. Wind turbine rotor blades are used to illustrate the principles discussed. These concepts are related to both International Electrotechnical Commission and Germanischer Lloyd design standards, and are covered using schematic representations of statistical load and material strength distributions. Wherever possible, interpretations of the partial safety factors are given with descriptions of their intended meaning. Under some circumstances, the authors` interpretations may be subjective. Next, the test-load factors are described in concept and then related to the design factors. Using technical arguments, it is shown that some of the design factors for both load and materials must be used in the test loading, but some should not be used. In addition, some test factors not used in the design may be necessary for an accurate test of the design. The results show that if the design assumptions do not clearly state the effects and uncertainties that are covered by the design`s partial safety factors, outside parties such as test labs or certification agencies could impose their own meaning on these factors.

Safety in the Chemical Laboratory: Tested Disposal Methods for Chemical Wastes from Academic Laboratories.

Science.gov (United States)

Armour, M. A.; And Others

1985-01-01

Describes procedures for disposing of dichromate cleaning solution, picric acid, organic azides, oxalic acid, chemical spills, and hydroperoxides in ethers and alkenes. These methods have been tested under laboratory conditions and are specific for individual chemicals rather than for groups of chemicals. (JN)

Geoelectrical Methods and Monitoring for Dam Safety Assessment, Republic of Korea

Science.gov (United States)

Lim, S. K.; Oldenburg, D.; Kang, S.; Song, S. H.

2016-12-01

Geoelectrical methods and monitoring to detect the seepage and internal erosion are essential for the safety assessment of earth dams. This work aims to develop improved methodologies to analyze the observed data and to monitor changes in seepage flow using direct current (DC) and self-potential (SP) methods. The seasonal variation of water level at dams causes a change in seepage and water saturation and hence alters the resistivity of the dam material. DC data are sensitive to water saturation and hence changes in saturation can be obtained by repeatedly measuring DC data. However, a more diagnostic parameter for safety assessment is fluid flow, and resistivity is only weakly coupled to that. Fortunately SP signals are directly related to fluid flow, and thus an SP survey has the potential to characterize fluid flow through the earth matrix. In Korea, the safety assessment of earth fill dams has been dealt by Korea Rural Community Corporation (KRC). Most of the dams are relatively old ( >50 years), hence assessing deterioration and corresponding seepage of those dams are crucial. In order to evaluate the engineering geological properties of the soil at earth dams in Korea, two boreholes in each dam were drilled to a bedrock depth that exceeds the height of the dam. A large set of field tests, including standard penetration tests (SPT) and in-situ permeability tests, were carried out along the boreholes. However, seepage paths in the dam is complex hence those limited measurements at a few points is not sufficient to delineate the zone of preferential seepage flow. For this, KRC developed permanent DC monitoring systems at a number of agricultural dams in Korea. The data were automatically collected every 6 hours. During the monitoring, the measurements of the water level at two boreholes were gathered at the same time. In this presentation we select an agricultural dam and delineate an anomalous leakage zone by inverting and interpreting time-lapse DC resistivity

Human reliability analysis methods for probabilistic safety assessment

International Nuclear Information System (INIS)

Pyy, P.

2000-11-01

Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

Safety Test Report for the SNF Dry Storage System

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Seo, K. S.; Lee, J. H.; Lee, J. C.; Choi, W. S

2008-11-15

This is technical report conducted by KAERI under the contract with NETEC for safety test for the PWR S/F dry storage system. Leak Test was performed after drop test and turn-over test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the dry storage system is maintained. In the seismic test, the moving of the model was measured at SRTH seismic response of 0.4 g and 0.8 g. Therefore the seismic test results can be used fully to the test data for verification of the seismic analysis. In the thermal test, the direction of the inlet and outlet of the air has no effect on the heat transfer performance. The passive heat removal system of the horizontal storage module was designed well.

Safety evaluation and regulation of chemicals. 2. Impact of regulations - improvement of methods

Energy Technology Data Exchange (ETDEWEB)

Homburger, F [ed.

1985-01-01

This volume assesses the impact of new scientific knowledge on the testing and regulation of chemicals, including food additives, drugs, cosmetics, pesticides, and other commercial substances. Apart from describing the newest tests, regulations, and risk assessment strategies, chapters reflect changes forced by both the growing need for cost containment and the mounting pressure to find alternatives to animal testing. Based on an international congress, the book also brings the advantage of diversity in the background and nationality of the authors, thus allowing a view of central problems according to the different interests of academics, industry scientists, government scientists, and regulators. The book opens with coverage of national and international regulations designed to prevent and control damage to human health and the environment. Topics range from basic problems of policy design and enforcement to the specific requirements for chemical regulation in developing countries. The next chapters cover new tests, systems, and assays used in in vivo safety testing. Readers will find a critical assessment of tests used to determine teratogenicity, mutagenicity, carcinogenicity, neurotoxicity and chemical lethality. Other topics include factors operating in the public perception of chemical hazards, guidelines for decision making in the management and regulation of risks, and future trends in the methodology of safety evaluation. The volume concludes with an overview of in vitro methods for testing hepatotoxicity. Several short-term in vitro test models and limited in vivo bioassays are presented and evaluated in terms of their capacity to substitute for long-term animal studies. Expert and thorough in its coverage, the book offers a wealth of technical and practical information for toxicologists, pharmacologists, industrial policy makers, and government regulators. (orig.). With 67 figs., 34 tabs.

An approach of sensitivity and uncertainty analyses methods installation in a safety calculation

International Nuclear Information System (INIS)

Pepin, G.; Sallaberry, C.

2003-01-01

Simulation of the migration in deep geological formations leads to solve convection-diffusion equations in porous media, associated with the computation of hydrogeologic flow. Different time-scales (simulation during 1 million years), scales of space, contrasts of properties in the calculation domain, are taken into account. This document deals more particularly with uncertainties on the input data of the model. These uncertainties are taken into account in total analysis with the use of uncertainty and sensitivity analysis. ANDRA (French national agency for the management of radioactive wastes) carries out studies on the treatment of input data uncertainties and their propagation in the models of safety, in order to be able to quantify the influence of input data uncertainties of the models on the various indicators of safety selected. The step taken by ANDRA consists initially of 2 studies undertaken in parallel: - the first consists of an international review of the choices retained by ANDRA foreign counterparts to carry out their uncertainty and sensitivity analysis, - the second relates to a review of the various methods being able to be used in sensitivity and uncertainty analysis in the context of ANDRA's safety calculations. Then, these studies are supplemented by a comparison of the principal methods on a test case which gathers all the specific constraints (physical, numerical and data-processing) of the problem studied by ANDRA

Methods for safety culture improvement

International Nuclear Information System (INIS)

Sivintsev, Yu.V.

1998-01-01

New IAEA publication concerning the problems of safety assurance covering different aspects beginning from terminology applied and up to concrete examples of well and poor safety culture development at nuclear facilities is discussed. The safety culture is defined as such set of characteristics and specific activities of institutions and individual persons which states that safety problems of a nuclear facility are given the attention determined by their importance as being of highest priority. The statements of the new document have recommended, not mandatory character. It is emphasized that the process of safety culture improvement at nuclear facilities should be integral component of management procedure, not a bolt on extra

Generic test platform for representative tests of safety I/C systems - 15546

International Nuclear Information System (INIS)

Fourestie, B.; Kuck, H.; Richter, J.; Rieche, S.; Waitz, M.

2015-01-01

In compliance with the IEC 61513 safety Instrumentation and Control (I/C) systems must be successfully validated in their final configuration prior to installation on site and commissioning. However the contingent need for modifications during system validation activities or subsequently during the commissioning phase may entail long and costly re-engineering of the I/C systems. With the view to ease these possible modifications, a Generic Test Platform has been developed by AREVA which allows combining a real I/C system subpart with an emulation server. This platform provides a faithful representation of the I/C System allowing crediting the validation test results carried out on this platform. (authors)

Operation safety of control systems. Principles and methods; Surete de fonctionnement des systemes de commande. Principes et methodes

Energy Technology Data Exchange (ETDEWEB)

Aubry, J.F. [Institut National Polytechnique, 54 - Nancy (France); Chatelet, E. [Universite de Technologie de Troyes, 10 (France)

2008-09-15

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.)

Standard test method for measurement of oxidation-reduction potential (ORP) of soil

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This test method covers a procedure and related test equipment for measuring oxidation-reduction potential (ORP) of soil samples removed from the ground. 1.2 The procedure in Section 9 is appropriate for field and laboratory measurements. 1.3 Accurate measurement of oxidation-reduction potential aids in the analysis of soil corrosivity and its impact on buried metallic structure corrosion rates. 1.4 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.5 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Inquiry relating to safety due to modification of usage of nuclear fuel material (establishment of waste safety testing facility) in Tokai Laboratory, Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1979-01-01

Application was made to the director of the Science and Technology Agency (STA) for the license relating to the modification of usage of nuclear fuel material (the establishment of waste safety testing facility) from the director of the Japan Atomic Energy Research Institute on November 30, 1978. After passing through the safety evaluation in the Nuclear Safety Bureau of STA, inquiry was conducted to the head of the Atomic Energy Safety Commission (AESC) on June 6, 1979, from the director of the STA. The head of AESC directed to conduct the safety examination to the head of the Nuclear Fuel Safety Examination Specialist Committee on June 7, 1979. The content of the modification of usage of nuclear fuel material is the establishment of waste safety testing facility to study and test the safety relating to the treatment and disposal of high level radioactive liquid wastes due to the reprocessing of spent fuel. As for the results of the safety examination, the siting of the waste safety testing facility which is located in the Tokai Laboratory, Japan Atomic Energy Research Institute (JAERI), and the test plan of the glass solidification of high level radioactive liquid are presented as the outline of the study plan. The building, main equipments including six cells, the isolation room and the glove box, the storage, and the disposal facilities for gas, liquid and solid wastes are explained as the outline of the facilities. Concerning the items from the viewpoint of safety, aseismatic design, slightly vacuum operation, shielding, decay heat removal, fire protection, explosion protection, criticality management, radiation management and environmental effect were evaluated, and the safety was confirmed. (Nakai, Y.)

Evaluation of Four Bedside Test Systems for Card Performance, Handling and Safety.

Science.gov (United States)

Giebel, Felix; Picker, Susanne M; Gathof, Birgit S

2008-01-01

SUMMARY: OBJECTIVE: Pretransfusion ABO compatibility testing is a simple and required precaution against ABO-incompatible transfusion, which is one of the greatest threats in transfusion medicine. While distinct agglutination is most important for correct test interpretation, protection against infectious diseases and ease of handling are crucial for accurate test performance. Therefore, the aim of this study was to evaluate differences in test card design, handling, and user safety. DESIGN: Four different bedside test cards with pre-applied antibodies were evaluated by 100 medical students using packed red blood cells of different ABO blood groups. Criteria of evaluation were: agglutination, labelling, handling, and safety regarding possible user injuries. Criteria were rated subjectively according to German school notes ranging from 1 = very good to 6 = very bad/insufficient. RESULTS: Overall, all cards received very good/good marks. The ABO blood group was identified correctly in all cases. Three cards (no. 1, no. 3, no. 4) received statistically significant (p labelling (1.5 vs. 2.2-2.4), handling (1.9-2.0 vs. 2.5), and user safety (2.5 vs. 3.4). Analysis of card self-explanation revealed no remarkable differences. CONCLUSION: Despite good performance of all card systems tested, the best results when including all criteria evaluated were obtained with card no. 4 (particularly concerning clear agglutination), followed by cards no. 2, no. 1, and no. 3.

Standard Test Method for Electronic Measurement for Hydrogen Embrittlement From Cadmium-Electroplating Processes

CERN Document Server

American Society for Testing and Materials. Philadelphia

1996-01-01

1.1 This test method covers an electronic hydrogen detection instrument procedure for measurement of plating permeability to hydrogen. This method measures a variable related to hydrogen absorbed by steel during plating and to the hydrogen permeability of the plate during post plate baking. A specific application of this method is controlling cadmium-plating processes in which the plate porosity relative to hydrogen is critical, such as cadmium on high-strength steel. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazard statement, see Section 8. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Safety analysis calculations for research and test reactors

Energy Technology Data Exchange (ETDEWEB)

Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

1983-08-01

The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

Advanced repair methods for enhanced reactor safety

International Nuclear Information System (INIS)

Kornfeldt, H.

1993-01-01

A few innovative concepts are described of the ABB Atom Service Division for repair and mitigation techniques for primary systems in nuclear power plants. The concepts are based on Shape Memory Alloy (SMA) technology. A basic feature of all methods is that welding and component replacement is being avoided and the radiation dose imposed on maintenance personnel reduced. The SMA-based repair methods give plant operators new ways to meet increased safety standards and rising maintenance costs. (Z.S.) 4 figs

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

Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

Directory of Open Access Journals (Sweden)

Zongzhi Li

2017-04-01

Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

Triangulation and the importance of establishing valid methods for food safety culture evaluation.

Science.gov (United States)

Jespersen, Lone; Wallace, Carol A

2017-10-01

The research evaluates maturity of food safety culture in five multi-national food companies using method triangulation, specifically self-assessment scale, performance documents, and semi-structured interviews. Weaknesses associated with each individual method are known but there are few studies in food safety where a method triangulation approach is used for both data collection and data analysis. Significantly, this research shows that individual results taken in isolation can lead to wrong conclusions, resulting in potentially failing tactics and wasted investments. However, by applying method triangulation and reviewing results from a range of culture measurement tools it is possible to better direct investments and interventions. The findings add to the food safety culture paradigm beyond a single evaluation of food safety culture using generic culture surveys. Copyright © 2017. Published by Elsevier Ltd.

Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

2006-07-01

The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

Safety and Function Test Report for the SWIFT Wind Turbine

Energy Technology Data Exchange (ETDEWEB)

Mendoza, I.; Hur, J.

2013-01-01

This test was conducted as part of the U.S. Department of Energy's (DOE) Independent Testing project. This project was established to help reduce the barriers of wind energy expansion by providing independent testing results for small turbines. Three turbines where selected for testing at the National Wind Technology Center (NWTC) as a part of round two of the Small Wind Turbine Independent Testing project. Safety and Function testing is one of up to 5 tests that may be performed on the turbines. Other tests include power performance, duration, noise, and power quality. The results of the testing will provide the manufacturers with reports that may be used for small wind turbine certification.

An abnormal situation modeling method to assist operators in safety-critical systems

International Nuclear Information System (INIS)