29 CFR 98.1000 - Respondent.

Science.gov (United States)

2010-07-01

... 29 Labor 1 2010-07-01 2010-07-01 true Respondent. 98.1000 Section 98.1000 Labor Office of the Secretary of Labor GOVERNMENTWIDE DEBARMENT AND SUSPENSION (NONPROCUREMENT) Definitions § 98.1000 Respondent. Respondent means a person against whom an agency has initiated a debarment or suspension action. ...

33 CFR 151.1000 - Purpose.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Purpose. 151.1000 Section 151.1000 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) POLLUTION... Transportation of Municipal and Commercial Waste § 151.1000 Purpose. The purpose of this subpart is to implement...

21 CFR 1000.1 - General.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false General. 1000.1 Section 1000.1 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) RADIOLOGICAL HEALTH GENERAL General Provisions § 1000.1 General. References in this subchapter J to regulatory sections of the...

25 CFR 1000.1 - Authority.

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false Authority. 1000.1 Section 1000.1 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL... § 1000.1 Authority. This part is prepared and issued by the Secretary of the Interior under the...

21 CFR 573.1000 - Verxite.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Verxite. 573.1000 Section 573.1000 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS... Listing § 573.1000 Verxite. The food additive verxite may be safely used in animal feed in accordance with...

25 CFR 11.1000 - Complaint.

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Complaint. 11.1000 Section 11.1000 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR LAW AND ORDER COURTS OF INDIAN OFFENSES AND LAW AND ORDER CODE Juvenile Offender Procedure § 11.1000 Complaint. A complaint must be filed by a law enforcement officer or...

30 CFR 250.1000 - General requirements.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 2 2010-07-01 2010-07-01 false General requirements. 250.1000 Section 250.1000... OPERATIONS IN THE OUTER CONTINENTAL SHELF Pipelines and Pipeline Rights-of-Way § 250.1000 General....1001, must meet the requirements in §§ 250.1000 through 250.1008. (2) A pipeline right-of-way grant...

7 CFR 1000.54 - Equivalent price.

Science.gov (United States)

2010-01-01

... 7 Agriculture 9 2010-01-01 2009-01-01 true Equivalent price. 1000.54 Section 1000.54 Agriculture... Prices § 1000.54 Equivalent price. If for any reason a price or pricing constituent required for computing the prices described in § 1000.50 is not available, the market administrator shall use a price or...

24 CFR 1000.28 - May a self-governance Indian tribe be exempted from the applicability of § 1000.26?

Science.gov (United States)

2010-04-01

... be exempted from the applicability of § 1000.26? 1000.28 Section 1000.28 Housing and Urban... ACTIVITIES General § 1000.28 May a self-governance Indian tribe be exempted from the applicability of § 1000... and systems meet or exceed the comparable requirements of § 1000.26. For purposes of this section, a...

AP1000. The PWR revisited

International Nuclear Information System (INIS)

Gaio, P.

2006-01-01

The distinguishing features of Westinghouse's AP1000 advanced passive pressurized water reactor are highlighted. In particular, the AP1000's passive safety features are described as well as their implications for simplifying the design, construction, and operation of this design compared to currently operating plants, and significantly increasing safety margins over current plants as well. The AP1000 design specifically incorporates the knowledge acquired from the substantial accumulation of power reactor operating experience and benefits from the application of the Probabilistic Risk Assessment in the design process itself. The AP1000 design has been certified by the US Nuclear Regulatory Commission under its new rules for licensing new nuclear plants, 10 CFR Part 52, and is the subject of six combined Construction and Operating License applications now being developed. Currently the AP1000 design is being assessed against the EUR Rev C requirements for new nuclear power plants in Europe. (author)

The Microstructures and Energy Dispersive Spectroscopy Analysis of a Hypoeutectoid Steels With 1% Cr

Directory of Open Access Journals (Sweden)

Rożniata E.

2013-12-01

Full Text Available W artykule zamieszczono wyniki badań mikrostruktury, analizy EDS i twardości stali podeutektoidalnych z 1% Cr imitujących składem chemicznym stale do ulepszania cieplnego. Badania dylatometryczne wykonano przy użyciu dylatometru L78R.I.T.A niemieckiej firmy LlNSElS. Za pomocą dyla- tometru rejestrowano zmiany wydłużenia (Al próbek o wymiarach >p 3X10mm w funkcji temperatury (T. Otrzymane krzywe nagrzewania posłużyły do precyzyjnego wyznaczenia temperatur krytycznych (punktów przełomowych dla badanych stali. Natomiast otrzymane krzywe chłodzenia różniczkowano, co pozwoliło precyzyjnie określić temperatury początków i końców poszczególnych przemian dla wykonania dwóch wykresów CTPc. Analizę składu chemicznego występujących faz w badanych stalach dla różnych szybkości chłodzenia wykonano przy użyciu mikrosondy elektronowej (mikroanalizator rentgenowski. W niniejszej pracy wykorzystano technikę analizy punktowej, liniowej oraz w obszarze o ustalonym polu. Po umieszczeniu próbek z badanych stali w komorze i osiągnięciu odpowiedniej próżni, wyszukano miejsca i dokonano analizy techniką EDS (Energy Dispersive Spectroscopy. Analizy EDS dokonano przy użyciu skaningowego mikroskopu transmisyjnego typu Nova NanoS EM 450. Opracowane wykresy CTPc zgodnie z klasyfikacją Wever’a i Rose’go są typu IV, co oznacza, że przemiany dyfuzyjne są rozdzielone zakresem trwałości przechłodzonego austenitu i mają kształt litery „C". Hartowność badanych stali jest zbliżona, jednak molibden działa zdecydowanie skuteczniej niż nikiel. Molibden zajmuje I-sze miejsce wśród skuteczności pierwiastków stopowych dla stali do niskiego odpuszczania, gdzie „tło" pierwiastków jest słabe. Dla obu badanych stali podeutektoidalnych analiza EDS wykazała wydzielanie się cementy tu stopowego na granicach ziaren. Chrom, jako pierwiastek terrytotwórczy dość silnie dyfunduje do granic ziaren. Widocznie jest to przy

Super cool X-1000 and Super cool Z-1000, two ice blockers, and their effect on vitrification/warming of mouse embryos.

Science.gov (United States)

Badrzadeh, H; Najmabadi, S; Paymani, R; Macaso, T; Azadbadi, Z; Ahmady, A

2010-07-01

To evaluate the survival and blastocyst formation rates of mouse embryos after vitrification/thaw process with different ice blocker media. We used X-1000 and Z-1000 separately and mixed using V-Kim, a closed vitrification system. Mouse embryos were vitrified using ethylene glycol based medium supplemented with Super cool X-1000 and/or Super cool Z-1000. Survival rates for the control, Super cool X-1000, Super cool Z-1000, and Super cool X-1000/Z-1000 groups were 74%, 72%, 68%, and 85% respectively, with no significant difference among experimental and control groups; however, a significantly higher survival rate was noticed in the Super cool X-1000/Z-1000 group when compared with the Super cool Z-1000 group. Blastocyst formation rates for the control, Super cool X-1000, Super cool Z-1000, and Super cool X-1000/Z-1000 groups were 71%, 66%, 65%, and 72% respectively. There was no significant difference in this rate among control and experimental groups. In a closed vitrification system, addition of ice blocker Super cool X-1000 to the vitrification solution containing Super cool Z-1000 may improve the embryo survival rate. We recommend combined ice blocker usage to optimize the vitrification outcome. Copyright (c) 2010 Elsevier Ireland Ltd. All rights reserved.

42 CFR 1000.10 - General definitions.

Science.gov (United States)

2010-10-01

... 42 Public Health 5 2010-10-01 2010-10-01 false General definitions. 1000.10 Section 1000.10 Public Health OFFICE OF INSPECTOR GENERAL-HEALTH CARE, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL PROVISIONS INTRODUCTION; GENERAL DEFINITIONS Definitions § 1000.10 General definitions. In this chapter...

7 CFR 1000.6 - Supply plant.

Science.gov (United States)

2010-01-01

... 7 Agriculture 9 2010-01-01 2009-01-01 true Supply plant. 1000.6 Section 1000.6 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements... Definitions § 1000.6 Supply plant. Supply plant means a plant approved by a duly constituted regulatory agency...

12 CFR 611.1000 - General authority.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 6 2010-01-01 2010-01-01 false General authority. 611.1000 Section 611.1000 Banks and Banking FARM CREDIT ADMINISTRATION FARM CREDIT SYSTEM ORGANIZATION Bank Mergers, Consolidations and Charter Amendments § 611.1000 General authority. (a) An amendment to a bank charter may relate...

AP1000, a nuclear central of advanced design; AP1000, una central nuclear de diseno avanzado

Energy Technology Data Exchange (ETDEWEB)

Hernandez M, N.; Viais J, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: nhm@nuclear.inin.mx

2005-07-01

The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

21 CFR 589.1000 - Gentian violet.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Gentian violet. 589.1000 Section 589.1000 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS... Substances Prohibited From Use in Animal Food or Feed § 589.1000 Gentian violet. The Food and Drug...

United States National Grid for New Mexico, UTM 12, (1000m X 1000m polygons )

Data.gov (United States)

Earth Data Analysis Center, University of New Mexico — This is a polygon feature data layer of United States National Grid (1000m x 1000m polygons ) constructed by the Center for Interdisciplinary Geospatial Information...

United States National Grid for New Mexico, UTM 13, (1000m X 1000m polygons )

Data.gov (United States)

Earth Data Analysis Center, University of New Mexico — This is a polygon feature data layer of United States National Grid (1000m x 1000m polygons ) constructed by the Center for Interdisciplinary Geospatial Information...

Westinghouse AP 1000 program status

International Nuclear Information System (INIS)

Doehnert, B.

2002-01-01

The project 1000 is presented and features are discussed in the paper. Design maturity is characterized by 1300 man-year / $400 million design and testing effort, more than 12 000 design documents completed; 3D computer model developed. It includes structures, equipment, small / large pipe, cable trays, ducts etc. Licensing Maturity is determined by a very thorough and complete NRC review of AP600; 110 man-year effort (NRC) over 6 years, $30 million; independent, confirmatory plant analysis; independent, confirmatory plant testing (ROSA, OSU); over 7400 questions answered, no open items; over 380 meeting with NRC, 43 meetings with ACRS. NRC Design Certification is issued in December 1999. Reasons for developing AP 1000 and design changes are presented. Economic analysis shows an expectation for payback within 20 years. AP1000 provides 75% power uprate for 15% increment in capital cost. AP1000 meets new plant economic targets in the near term

AP1000{sup TM} plant modularization

Energy Technology Data Exchange (ETDEWEB)

Cantarero L, C.; Demetri, K. J. [Westinghouse Electric Co., 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Quintero C, F. P., E-mail: cantarc@westinghouse.com [Westinghouse Electric Spain, Padilla 17, 28006 Madrid (Spain)

2016-09-15

The AP1000{sup TM} plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

AP1000"T"M plant modularization

International Nuclear Information System (INIS)

Cantarero L, C.; Demetri, K. J.; Quintero C, F. P.

2016-09-01

The AP1000"T"M plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

Generation of live offspring from vitrified embryos with synthetic polymers SuperCool X-1000 and SuperCool Z-1000.

Science.gov (United States)

Marco-Jimenez, F; Jimenez-Trigos, E; Lavara, R; Vicente, J S

2014-01-01

Ice growth and recrystallisation are considered important factors in determining vitrification outcomes. Synthetic polymers inhibit ice formation during cooling or warming of the vitrification process. The aim of this study was to assess the effect of adding commercially available synthetic polymers SuperCool X-1000 and SuperCool Z-1000 to vitrification media on in vivo development competence of rabbit embryos. Four hundred and thirty morphologically normal embryos recovered at 72 h of gestation were used. The vitrification media contained 20% dimethyl sulphoxide and 20% ethylene glycol, either alone or in combination with 1% of SuperCool X-1000 and 1% SuperCool. Our results show that embryos can be successfully vitrified using SuperCool X-1000 and SuperCool Z-1000 and when embryos are transferred, live offspring can be successfully produced. In conclusion, our results demonstrated that we succeeded for the first time in obtaining live offspring after vitrification of embryos using SuperCool X-1000 and SuperCool Z-1000 polymers.

VVER-1000 backfitting programs

International Nuclear Information System (INIS)

Zabka, H.; Milhem, J.L.

1998-01-01

Russia, Ukraine, and Bulgaria have nineteen nuclear generating units of the VVER-1000/V-320 (1000 MWe PWR) type in operation. Most of these plants were built in the eighties. Their design is based on Soviet standards of the seventies. In the early eighties and, in particular, after the Chernobyl accident, new safety principles and supplementary specific standards were introduced. However, they were taken into account only to a limited extent in the design and construction of the VVER-1000/V-320 plants. A number of nuclear power plants, whose construction was stopped after the political changes in the countries of the former USSR, now are to be completed with the financial assistance of the Commission of the European Union and other Western organizations, respectively. This Western support is dependent on the condition that these plants attain a level of engineered safeguards comparable to that of PWR plants currently in operation in Western Europe. (orig.) [de

7 CFR 1940.1000 - OMB control number.

Science.gov (United States)

2010-01-01

... 7 Agriculture 13 2010-01-01 2009-01-01 true OMB control number. 1940.1000 Section 1940.1000 Agriculture Regulations of the Department of Agriculture (Continued) RURAL HOUSING SERVICE, RURAL BUSINESS....1000 OMB control number. The collection of information requirements contained in this regulation has...

26 CFR 1.998-1.1000 - [Reserved

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 10 2010-04-01 2010-04-01 false [Reserved] 1.998-1.1000 Section 1.998-1.1000 Internal Revenue INTERNAL REVENUE SERVICE, DEPARTMENT OF THE TREASURY (CONTINUED) INCOME TAX (CONTINUED) INCOME TAXES Domestic International Sales Corporations §§ 1.998-1.1000 [Reserved] ...

7 CFR 4284.1000 - OMB control number.

Science.gov (United States)

2010-01-01

... 7 Agriculture 15 2010-01-01 2010-01-01 false OMB control number. 4284.1000 Section 4284.1000 Agriculture Regulations of the Department of Agriculture (Continued) RURAL BUSINESS-COOPERATIVE SERVICE AND RURAL UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE GRANTS Value-Added Producer Grants § 4284.1000 OMB...

OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis

International Nuclear Information System (INIS)

Ivanov, B.; Ivanov, K.; Aniel, S.; Royer, E.; Kolev, N.; Groudev, P.

2004-01-01

The present paper describes the two phases of the OECD/DOE/CEA VVER-1000 coolant transient benchmark labeled as V1000CT. This benchmark is based on a data from the Bulgarian Kozloduy NPP Unit 6. The first phase of the benchmark was designed for the purpose of assessing neutron kinetics and thermal-hydraulic modeling for a VVER-1000 reactor, and specifically for their use in analyzing reactivity transients in a VVER-1000 reactor. Most of the results of Phase 1 will be compared against experimental data and the rest of the results will be used for code-to-code comparison. The second phase of the benchmark is planned for evaluation and improvement of the mixing computational models. Code-to-code and code-to-data comparisons will be done based on data of a mixing experiment conducted at Kozloduy-6. Main steam line break will be also analyzed in the second phase of the V1000CT benchmark. The results from it will be used for code-to-code comparison. The benchmark team has been involved in analyzing different aspects and performing sensitivity studies of the different benchmark exercises. The paper presents a comparison of selected results, obtained with two different system thermal-hydraulics codes, with the plant data for the Exercise 1 of Phase 1 of the benchmark as well as some results for Exercises 2 and 3. Overall, this benchmark has been well accepted internationally, with many organizations representing 11 countries participating in the first phase of the benchmark. (authors)

7 CFR 1000.14 - Other source milk.

Science.gov (United States)

2010-01-01

... 7 Agriculture 9 2010-01-01 2009-01-01 true Other source milk. 1000.14 Section 1000.14 Agriculture... and Orders; Milk), DEPARTMENT OF AGRICULTURE GENERAL PROVISIONS OF FEDERAL MILK MARKETING ORDERS Definitions § 1000.14 Other source milk. Other source milk means all skim milk and butterfat contained in or...

25 CFR 39.1000 - Purpose and scope.

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Purpose and scope. 39.1000 Section 39.1000 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR EDUCATION THE INDIAN SCHOOL EQUALIZATION PROGRAM Administrative Cost Formula § 39.1000 Purpose and scope. The purpose of this subpart is to provide funds at the...

27 CFR 19.1000 - Reconsignment in transit.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Reconsignment in transit. 19.1000 Section 19.1000 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE..., Withdrawals and Transfers § 19.1000 Reconsignment in transit. When, prior to or on arrival at the premises of...

24 CFR 35.1000 - Purpose and applicability.

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Purpose and applicability. 35.1000 Section 35.1000 Housing and Urban Development Office of the Secretary, Department of Housing and Urban..., Support Services, or Operation § 35.1000 Purpose and applicability. (a) The purpose of this subpart K is...

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the First Workshop (V1000-CT1)

International Nuclear Information System (INIS)

2003-01-01

The first workshop for the VVER-1000 Coolant Transient Benchmark TT Benchmark was hosted by the Commissariat a l'Energie Atomique, Centre d'Etudes de Saclay, France. The V1000CT benchmark defines standard problems for validation of coupled three-dimensional (3-D) neutron-kinetics/system thermal-hydraulics codes for application to Soviet-designed VVER-1000 reactors using actual plant data without any scaling. The overall objective is to access computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transient simulations in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 - simulation of the switching on of one main coolant pump (MCP) while the other three MCP are in operation, and V1000CT- 2 - calculation of coolant mixing tests and Main Steam Line Break (MSLB) scenario. Further background information on this benchmark can be found at the OECD/NEA benchmark web site . The purpose of the first workshop was to review the benchmark activities after the Starter Meeting held last year in Dresden, Germany: to discuss the participants' feedback and modifications introduced in the Benchmark Specifications on Phase 1; to present and to discuss modelling issues and preliminary results from the three exercises of Phase 1; to discuss the modelling issues of Exercise 1 of Phase 2; and to define work plan and schedule in order to complete the two phases

ACR-1000: Operator - based development

International Nuclear Information System (INIS)

Shalaby, B.; Alizadeh, A.

2007-01-01

Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU * reactors to establish Generation III + Advanced CANDU Reactor T M (ACR T M) technology. The ACR-1000 T M nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability. The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution. Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDU T M technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants. As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of

7 CFR 1000.86 - Deduction for marketing services.

Science.gov (United States)

2010-01-01

... 7 Agriculture 9 2010-01-01 2009-01-01 true Deduction for marketing services. 1000.86 Section 1000... SERVICE (Marketing Agreements and Orders; Milk), DEPARTMENT OF AGRICULTURE GENERAL PROVISIONS OF FEDERAL MILK MARKETING ORDERS Administrative Assessment and Marketing Service Deduction § 1000.86 Deduction for...

EP 1000 -The European Passive Plant

International Nuclear Information System (INIS)

Cummins, Ed; Oyarzabal, Mariano; Saiu, Gianfranco

1998-01-01

A group of European utilities, along with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT) initiated a program to evaluate Westinghouse passive nuclear plant technology for application in Europe. The European utility group consisted of: Agrupacion electrica para al Desarrollo Technologico Nuclear (DTN), Spain; Electricite de France; ENEL, SpA., Italy; IVO Power Engineering, Ltd., Finland; Scottish Nuclear Limited (acting for itself on behalf of Nuclear Electric plc, U.K.; Tractebel Energy Engineering, Belgium; UAK (represented by NOK-Beznau), Switzerland; and Vattenfall AB, Ringhals, Sweden. The European Passive Plant (EPP) program, which began in 1994, is an evaluation of the Westinghouse 600 MWe AP 600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs in meeting the European Utility Requirements (EUR), and where necessary, modifying the design to achieve compliance. Phase 1 or the EPP program was completed and included the two major tasks of evaluating the effect of the EUR on the Westinghouse nuclear island and developing the EP 1000, a 1000 MWe passive plant reference design that conforms to the EUR and would be licensable in Europe. The EP 1000 closely follows the Westinghouse SPWR design for safety systems and containment and the AP 600 design for auxiliary systems. It also includes features that where required to meet the EUR and key European licensing requirements. The primary circuit of the EP 1000 retains most of the general features of the current-day designs, but some evolutionary features to enhance reliability, simplicity of operation, ease of maintenance, and plant safety have been incorporated into the design. The core, reactor vessel, and reactor internals of the EP 1000 are similar to those of currently operating Westinghouse PWR plants, but several new features are included to enhance the performance characteristics. The basic EP 1000 safety philosophy is based on use of inherent

21 CFR 892.1000 - Magnetic resonance diagnostic device.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Magnetic resonance diagnostic device. 892.1000 Section 892.1000 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1000 Magnetic resonance diagnostic...

42 CFR 1000.20 - Definitions specific to Medicare.

Science.gov (United States)

2010-10-01

... 42 Public Health 5 2010-10-01 2010-10-01 false Definitions specific to Medicare. 1000.20 Section 1000.20 Public Health OFFICE OF INSPECTOR GENERAL-HEALTH CARE, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL PROVISIONS INTRODUCTION; GENERAL DEFINITIONS Definitions § 1000.20 Definitions specific to...

13 CFR 120.1000 - Risk-Based Lender Oversight.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Risk-Based Lender Oversight. 120.1000 Section 120.1000 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Risk-Based Lender Oversight Supervision § 120.1000 Risk-Based Lender Oversight. (a) Risk-Based Lender...

Scandinavia 1000-1750

DEFF Research Database (Denmark)

Rasmussen, Carsten Porskrog; Myking, John Ragnar

2010-01-01

The article describes the distribution of property and power in rural Scandinavia (Denmark, Sweden and Norway) a.D. 1000-1750. It covers the distribution of property and structure of manors and holdings. It also looks at the structure of the peasantry, village communities, and the use of communal...

ACR-1000 design provisions for severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

Poučevanje biologije z uporabo elektronskega učbenika

OpenAIRE

Arh, Ana

2017-01-01

Z vpeljevanjem rabe tabličnih računalnikov v šole so se začele pojavljati tudi raziskave učinka digitalnih vsebin na znanje in stališča učencev. Pozitivni vidiki se kažejo pri motivaciji učencev za delo, hitrejšem in enostavnejšem dostopu do podatkov, enostavnem prenašanju po prostoru, učinkovitejšem zapisovanju, boljšem sodelovanju med učenci ter boljšem sodelovanju učencev z učiteljem. Vsi ti vidiki se nanašajo predvsem na stališča učencev do tabličnih računalnikov, zelo malo pa je podatkov...

10 CFR 2.1000 - Scope of subpart J.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Scope of subpart J. 2.1000 Section 2.1000 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Procedures... Geologic Repository § 2.1000 Scope of subpart J. The rules in this subpart, together with the rules in...

Steam generator design requirements for ACR-1000

International Nuclear Information System (INIS)

Subash, S.; Hau, K.

2006-01-01

Atomic Energy of Canada Limited (AECL) has developed the ACR-1000 (Advanced CANDU Reactor-1000 ) to meet market expectations for enhanced safety of plant operation, high capacity factor, low operating cost, increased operating life, simple component replacement, reduced capital cost, and shorter construction schedule. The ACR-1000 design is based on the use of horizontal fuel channels surrounded by a heavy water moderator, the same feature as in all CANDU reactors. The major innovation in the ACR-1000 is the use of low enriched uranium fuel, and light water as the coolant, which circulates in the fuel channels. This results in a compact reactor core design and a reduction of heavy water inventory, both contributing to a significant decrease in capital cost per MWe produced. The ACR-1000 plant is a two-unit, integrated plant with each unit having a nominal gross output of about 1165 MWe with a net output of approximately 1085 MWe. The plant design is adaptable to a single unit configuration, if required. This paper focuses on the technical considerations that went into developing some of the important design requirements for the steam generators for the ACR-1000 plant and how these requirements are specified in the Technical Specification, which is the governing document for the steam generator (SG) detail design. Layout of these SGs in the plant is briefly described and their impacts on the SG design. (author)

AP1000 design and construction integration

International Nuclear Information System (INIS)

Winters, James W.; Clelland, Jill A.

2004-01-01

Construction costs of commercial nuclear generating plants must be reduced in order to expand the future use of nuclear energy. Two of the drivers of plant construction costs are the cost of financing during the construction duration and the substantial amount of skilled craft labor hours needed on site during construction. The application of information technology (IT) has been used to understand and reduce both of these drivers by establishing parallel construction paths using modules and integrating construction sequence review into the design process. In a program sponsored by EPRI, Westinghouse has modeled the construction of AP1000 in '4D' to show its viability, to improve its logic, to improve the plant design for constructibility and overall to reduce time and risk in the construction schedule. The design of most of AP1000 was constrained to be a duplicate of AP600 except where components required expansion for the higher power level. As a result, the construction schedule for AP1000 is as mature and as robust as that for AP600. Two areas important to the construction of AP1000 did require some design work because they could not remain the same as AP1000. First, the turbine building had to be redesigned to accommodate the larger turbine and its support systems. Again, as much of the AP600 design and philosophy as possible was retained. The building required enlargement and the basemat, foundations, steel structure and structural modules required modification. As concrete, steel, and equipment were defined by the designers, they were matched to the original AP600 turbine building schedule. This forced designers to assemble files to be consistent with building assembly activities and to think about constructibility as they defined the final design. Second, the reinforcement structure within the concrete under and supporting the containment vessel required detail design. Westinghouse was fortunate to have the constructor Obayashi of Japan recommend a detailed

ACR-1000: Enhanced response to severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

AP1000, a nuclear central of advanced design

International Nuclear Information System (INIS)

Hernandez M, N.; Viais J, J.

2005-01-01

The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

77 FR 64379 - Proposed Collection; Comment Request for Form 1000

Science.gov (United States)

2012-10-19

... 1000 AGENCY: Internal Revenue Service (IRS), Treasury. ACTION: Notice and request for comments. SUMMARY... 1000, Ownership Certificate. DATES: Written comments should be received on or before December 18, 2012...: Ownership Certificate. OMB Number: 1545-0054. Form Number: 1000. Abstract: Form 1000 is used by citizens...

16 CFR 1000.26 - Directorate for Epidemiology.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Directorate for Epidemiology. 1000.26... AND FUNCTIONS § 1000.26 Directorate for Epidemiology. The Directorate for Epidemiology, managed by the Associate Executive Director for Epidemiology, is responsible for the collection and analysis of data on...

European utility requirements (EUR) volume 3 assessment for AP1000

International Nuclear Information System (INIS)

Saiu, G.; Demetri, K.J.

2005-01-01

The EUR (European Utility Requirements) Volume 3 is intended to report the Plant Description, the Compliance Assessment to EUR Volumes 1 and 2, and finally, the Specific Requirements for each specific Nuclear Power Plant Design considered by the EUR. Five subsets of EUR Volume 3, based on EUR Revision B, are already published; all of which are next generation plant designs being developed for Europe beyond 2000. They include : 1) EP1000 - Passive Pressurized Light Water Reactor (3-Loop, 1000 MWe) 2) EPR - Evolutionary Pressurized Light Water Reactor (1500 MWe) 3) BWR90/90+ - Evolutionary Boiling Water Reactor (1400 MWe) 4) ABWR - Evolutionary Boiling Water Reactor (1400 MWe) 5) SWR 1000 - Boiling Water Reactor With Passive Features (1000 MWe) In addition, the following subsets are currently being developed: 1) AP1000 - Passive Pressurized Light Water Reactor (2-Loop, 1117 MWe) 2) VVER AES 92 - Pressurized Water Reactor With Passive Features (1000 MWe) The purpose of this paper is to provide an overview of the program, which started in January 2004 with the EUR group to prepare an EUR Volume 3 Subset for the AP1000 nuclear plant design. The AP1000 EUR compliance assessment, to be performed against EUR Revision C requirements, is an important step for the evaluation of the AP1000 design for application in Europe. The AP1000 compliance assessment is making full use of AP1000 licensing documentation, EPP Phase 2 design activities and EP1000 EUR detailed compliance assessment. As of today, nearly all of the EUR Chapters have been discussed within the EUR Coordination Group. Based on the results of the compliance assessment, it can be stated that the AP1000 design shows a good level of compliance with the EUR Revision C requirements. Nevertheless, the compliance assessment has highlighted areas for where the AP1000 plant deviates from the EUR. The EPP design group has selected the most significant ones for performing detailed studies to quantify the degree of compliance

AP1000 - the new standard for nuclear power

International Nuclear Information System (INIS)

Lipman, Daniel S.

2006-01-01

Full text of publication follows: The AP1000 is the only Generation III+ reactor to receive Final Design Approval (FDA) from the Nuclear Regulatory Commission, and is expected to receive its Design Certification by the end of the year. Building on the proven features of current generation nuclear plants, the AP1000 combines experience with innovation into a design that surpasses current standards of safety and reliability. Use of passive safety features results in a simpler and more compact design that enhances safety, simplifies O and M requirements, and reduces capital and operating costs. At 1117 Mwe, the AP1000 is well suited for almost any grid system and will be fully competitive with combined-cycle gas and comparable fossil fuel plants. The AP1000 is ready to help launch a renaissance in new nuclear plant construction throughout the world. Maturity of Design: In excess of 1300 man-years and $400 million in development funding have been expended on the AP1000. It has undergone extensive, part scale testing at the system, sub-system and component level, in addition to a series of part scale integrated tests. The AP1000 is the most analyzed of the next generation reactors. Simplicity of Design/Economics: The AP1000 uses simplified and innovative passive safety systems to an unprecedented extent. Simplified passive safety systems provide reliable operation, reduced capital costs, and enhanced plant safety with large plant operating margins. The AP1000 features improved reliability through simplicity rather than addition of redundant safety trains. This simpler design is easier and less costly to operate and maintain than larger, more complex plants, while less equipment and smaller buildings translate into lower capital costs and shorter construction durations. After construction, economic benefit will be found in reduced operating and maintenance costs, largely due to reduced operating and maintenance staffing requirements. Construction aspects

24 CFR 1000.16 - What labor standards are applicable?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What labor standards are applicable? 1000.16 Section 1000.16 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.16 What labor standards are...

24 CFR 1000.18 - What environmental review requirements apply?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What environmental review requirements apply? 1000.18 Section 1000.18 Housing and Urban Development Regulations Relating to Housing and... HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.18 What environmental...

24 CFR 1000.12 - What nondiscrimination requirements are applicable?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What nondiscrimination requirements are applicable? 1000.12 Section 1000.12 Housing and Urban Development Regulations Relating to Housing..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.12 What...

40 CFR 60.1000 - What does this subpart do?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 6 2010-07-01 2010-07-01 false What does this subpart do? 60.1000 Section 60.1000 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS... Modification or Reconstruction is Commenced After June 6, 2001 Introduction § 60.1000 What does this subpart do...

20 CFR 408.1000 - What is this subpart about?

Science.gov (United States)

2010-04-01

... 20 Employees' Benefits 2 2010-04-01 2010-04-01 false What is this subpart about? 408.1000 Section 408.1000 Employees' Benefits SOCIAL SECURITY ADMINISTRATION SPECIAL BENEFITS FOR CERTAIN WORLD WAR II... Determinations § 408.1000 What is this subpart about? (a) Explanation of the administrative review process. This...

16 CFR 1000.11 - Delegation of functions.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Delegation of functions. 1000.11 Section... FUNCTIONS § 1000.11 Delegation of functions. Section 27(b)(9) of the Consumer Product Safety Act (15 U.S.C... to issue subpoenas, to any officer or employee of the Commission. Delegations are documented in the...

25 CFR 1000.354 - What is a trust evaluation?

Science.gov (United States)

2010-04-01

... that the functions are performed in accordance with trust standards as defined by Federal law. Trust... 25 Indians 2 2010-04-01 2010-04-01 false What is a trust evaluation? 1000.354 Section 1000.354... Trust Evaluation Review Annual Trust Evaluations § 1000.354 What is a trust evaluation? A trust...

WWER-1000 Burnup Credit Benchmark (CB5)

International Nuclear Information System (INIS)

Manolova, M.A.

2002-01-01

In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)

Key developments of the EP1000 design

International Nuclear Information System (INIS)

Noviello, L.

1999-01-01

In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The Phase I of the European Passive Plant program involved the evaluation of the Westinghouse 600 MWe AP600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs against the European Utility Requirements (EUR), and when necessary, the investigation of possible modifications to achieve compliance with the EUR. In Phase 1 of the program, which has been completed in 1996, the following major tasks were accomplished: The impacts of the European Utility Requirements (EUR) on the Westinghouse nuclear island design were evaluated. A 1000 MWe passive plant reference design (EP1000) was developed which conforms to the EUR and is expected to be licensable in Europe. With respect to the NSSS and containment, the EP1000 reference design closely follows those of the Westinghouse SPWR design, while the AP600 design has been taken as the basis for the design of the auxiliary systems. Extensive design and testing efforts have been made for the AP600 and SPWR during the respective multi-year programs. While the results of these programs have been and will continue to be utilised, at the maximum extent, to minimise the work to be performed on the EP1000 design, the compliance with EUR is a key design requirement for the EP1000 The ultimate objective of Phase 2 of the program is to develop design details and perform supporting analyses to produce a Safety Case Report (SCR) for submittal to European Safety Authorities. The first part of Phase 2, hereafter referred as Phase 2A, started at the beginning of 1997 and will be completed at the end of 1998. Scope of this phase of the program is to develop the design modifications of important systems and structures so to comply with the

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)

International Nuclear Information System (INIS)

2005-01-01

The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications

Radiation embrittlement of WWER-1000 reactor vessel steels

International Nuclear Information System (INIS)

Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

2001-01-01

Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

40 CFR 158.1000 - Applicator exposure-general requirements.

Science.gov (United States)

2010-07-01

... requirements. 158.1000 Section 158.1000 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED... Occupational Safety and Health Administration (OSHA), provide adequate protection from risk under FIFRA for a... and the applicator activity of concern, including formulation type, application methods and rates...

The Soviet RBMK-1000 containment system

International Nuclear Information System (INIS)

Joosten, J.K.

1988-01-01

Following the accident in April, 1986, considerable attention was focused on the failure of the containment at the Chernobyl RBMK-1000 nuclear power plant. Conflicting statements arose regarding the nature of the plant's containment system primarily because of terminology differences, translation difficulties and lack of reliable information. This article, based on reports and briefings by the Soviet delegation, during the post-accident review meetings in Vienna and prior publications is intended to clarify perceptions of the Soviet RMBK-1000 nuclear power plant containment system design, and its relevance to containment management concepts. (author)

A Probability Analysis of the Generating Cost for APR1000+

Energy Technology Data Exchange (ETDEWEB)

Ha, Gag-Hyeon; Kim, Dae-Hun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, cutting CO{sub 2} emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the APR1000+ development is to ensure export competitiveness in the developing countries in the Middle East and Southeast Asia. To that end, APR1000+(1,000MWe, 3.5 generation) will be developed based on APR+ (1,500MWe, 3.5 generation). And comparing to OPR1000(Korean Standard Nuclear power Plant, 2.5 generation), APR1000+ have many design features such as the 60 year design life time, comprehensive site requirement of 0.3g seismic design, stability improvement, operability improvement and provisions for severe accidents. In this simulation, the results of generating cost for APR1000+ preliminary conceptual design using a probability method was shown to be 48.37 ~ 74.22 won/kWh(median value 56.51 won/kWh). Those of OPR1000 was 42.08 ~ 61.77 won/kWh(median value 48.63 won/kWh). APR1000+ has -16.2% cost advantage over OPR1000 nuclear power plant. The main reason of this results is due to adding several safety designs.

AP1000 plant construction in China: Ansaldo Nucleare contribution

International Nuclear Information System (INIS)

Frogheri, Monica; Saiu, Gianfranco

2009-01-01

On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract includes: the design of the steel containment vessel, preparation of construction and fabrication, specifications, design and supply of SCV mechanical penetrations, air locks and equipment hatches. Moreover, Ansaldo Nucleare is in charge of the final design of the AP1000 PRHR-HX and together with Mangiarotti Nuclear will supply the component for the Sanmen Unit 1 NPP. The paper presents an overview of the design and manufacturing activities performed by Ansaldo Nucleare and its partners for the AP1000 plant in China. (authors)

AP1000R licensing and deployment in the United States

International Nuclear Information System (INIS)

Jordan, R. P.; Russ, P. A.; Filiak, P. P.; Castiglione, L. L.

2012-01-01

In recent years, both domestic and foreign utilities have turned to the standardized Westinghouse AP1000 plant design in satisfying their near - and long-term - sustainable energy needs. As direct support to these actions, licensing the AP1000 design has played a significant role by providing one of the fundamental bases in clearing regulatory hurdles leading to the start of new plant construction. Within the U.S. alone, Westinghouse AP1000 licensing activities have reached unprecedented milestones with the approvals of both AP1000 Design Certification and Southern Company's combined construction permit and operating license (COL) application directly supporting the construction of two new nuclear plants in Georgia. Further COL application approvals are immediately pending for an additional two AP1000 plants in South Carolina. And, across the U.S. nuclear industry spectrum, there are 10 other COL applications under regulatory review representing some 16 new plants at 10 sites. In total, these actions represent the first wave of new plant licensing under the regulatory approval process since 1978. Fundamental to the Nuclear Regulatory Commission's AP1000 Design Certification is the formal recognition of the AP1000 passive safety design through regulatory acceptance rulemaking. Through recognition and deployment of the AP1000 Design Certification, the utility licensee / operator of this reactor design are now offered an opportunity to use a simplified 'one-step' combined license process, thereby managing substantial back-end construction schedule risk from regulatory and intervention delays. Application of this regulatory philosophy represents both acceptance and encouragement of standardized reactor designs like the AP1000. With the recent AP1000 Design Certification and utility COL acceptances, the fundamental licensing processes of this philosophy have successfully proven the attainment of significant milestones with the next stage licensing actions directed

16 CFR 1000.29 - Directorate for Engineering Sciences.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Directorate for Engineering Sciences. 1000... ORGANIZATION AND FUNCTIONS § 1000.29 Directorate for Engineering Sciences. The Directorate for Engineering Sciences, which is managed by the Associate Executive Director for Engineering Sciences, is responsible for...

Westinghouse AP1000 advanced passive plant: design features and benefits

International Nuclear Information System (INIS)

Walls, S.J.; Cummins, W.E.

2003-01-01

The Westinghouse AP1000 Program is aimed at implementing the AP1000 plant to provide a further major improvement in plant economics while maintaining the passive safety advantages established by the AP600. An objective is to retain to the maximum extent possible the plant design of the AP600 so as to retain the licensing basis, cost estimate, construction schedule, modularization scheme, and the detailed design from the AP600 program. Westinghouse and the US Nuclear Regulatory Commission staff have embarked on a program to complete Design Certification for the AP1000 by 2004. A pre-certification review phase was completed in March 2002 and was successful in establishing the applicability of the AP600 test program and AP600 safety analysis codes to the AP1000 Design Certification. On March 28, 2002, Westinghouse submitted to US NRC the AP1000 Design Control Document and Probabilistic Risk Assessment, thereby initiating the formal design certification review process. The results presented in these documents verify the safety performance of the API 000 and conformance with US NRC licensing requirements. Plans are being developed for implementation of a series of AP1000 plants in the US. Key factors in this planning are the economics of AP1000, and the associated business model for licensing, constructing and operating these new plants. Similarly plans are being developed to get the AP1000 design reviewed for use in the UK. Part of this planning has been to examine the AP1000 design relative to anticipated UK safety and licensing issues. (author)

24 CFR 1000.102 - What are eligible affordable housing activities?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What are eligible affordable housing activities? 1000.102 Section 1000.102 Housing and Urban Development Regulations Relating to... § 1000.102 What are eligible affordable housing activities? Eligible affordable housing activities are...

16 CFR 1000.27 - Directorate for Health Sciences.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Directorate for Health Sciences. 1000.27... AND FUNCTIONS § 1000.27 Directorate for Health Sciences. The Directorate for Health Sciences is managed by the Associate Executive Director for Health Sciences and is responsible for reviewing and...

7 CFR 1000.19 - Commercial food processing establishment.

Science.gov (United States)

2010-01-01

... 7 Agriculture 9 2010-01-01 2009-01-01 true Commercial food processing establishment. 1000.19 Section 1000.19 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements and Orders; Milk), DEPARTMENT OF AGRICULTURE GENERAL PROVISIONS OF FEDERAL MILK MARKETING ORDERS Definitions §...

24 CFR 1000.101 - What is affordable housing?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What is affordable housing? 1000.101 Section 1000.101 Housing and Urban Development Regulations Relating to Housing and Urban Development (Continued) OFFICE OF ASSISTANT SECRETARY FOR PUBLIC AND INDIAN HOUSING, DEPARTMENT OF HOUSING AND...

EP1000 passive plant description

International Nuclear Information System (INIS)

Saiu, G.

1999-01-01

In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In Phase I of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) will be completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. The second part, 'Phase 2B', includes both the analyses and evaluations required to demonstrate the adequacy of the design, and to support the preparation of Safety Case Report. The second part of Phase 2 of the program will start at the beginning of 1999 and will be completed in the 2001. Incorporation of the EUR has been a key design requirement for the EP1000 from the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. This paper integrates and updates the plant description reported in the IAEA TECDOC-968. The most significant developments of the EP1000 plant design during Phase 2A of the EPP program are described and reference is made to the key design requirements set by the EUR Rev. B document. (author)

AccountAbility 1000: a new social standard for building sustainability

NARCIS (Netherlands)

Beckett, R.; Jonker, J.

2002-01-01

The pressing need to build common frameworks to redefine the performance and sustainability of organisations, has led to the development of a new standard, AccountAbility 1000 (AA1000). AA1000 is a quality framework that aims to make clear how principles of accountability and sustainability are

30 CFR 285.1000 - What activities does this subpart regulate?

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 2 2010-07-01 2010-07-01 false What activities does this subpart regulate? 285.1000 Section 285.1000 Mineral Resources MINERALS MANAGEMENT SERVICE, DEPARTMENT OF THE INTERIOR... Activities § 285.1000 What activities does this subpart regulate? (a) This subpart provides the general...

24 CFR 1000.46 - Do drug-free workplace requirements apply?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Do drug-free workplace requirements apply? 1000.46 Section 1000.46 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.46 Do drug-free workplace...

24 CFR 1000.38 - What flood insurance requirements are applicable?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What flood insurance requirements are applicable? 1000.38 Section 1000.38 Housing and Urban Development Regulations Relating to Housing..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.38 What flood...

24 CFR 1000.4 - What are the objectives of NAHASDA?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What are the objectives of NAHASDA? 1000.4 Section 1000.4 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.4 What are the objectives of NAHASDA...

Essence and characteristics of the Westinghouse technology AP1000

International Nuclear Information System (INIS)

Llovet, Ricardo

2014-01-01

The AP1000 nuclear power plant can place the reactor in a Safe Shutdown Condition within the first 72 hours of a Station Blackout, without the use of AC power or operator action •With some operator action after 3 days, the AP1000 nuclear power plant continues to maintain reactor core cooling and Spent Fuel Pool cooling indefinitely •The AP1000 nuclear power plant has superior coping capabilities as well as significantly reduced risk for core damage

Learning through delivery, Westinghouse AP1000 plant construction

International Nuclear Information System (INIS)

Gorgemans, J.; Hinman, R.D.; Steuck, C.M.; Greco, P.L.

2014-01-01

The AP1000 plant, which is a 1100 MWe class pressurized water reactor with passive safety features, is designed around a conventional 2 loop, 2 steam generator primary system configuration with 2 hot legs, 4 reactor coolant pumps directly mounted in the steam generator lower head and 4 cold legs. A particular feature of AP1000 is its modular construction to minimize the time and cost of construction. Modular construction allows activities to be run in parallel, it allows more activities to be performed in a controlled factory instead of in the field, and it provides a better level of quality. The AP1000 plant design includes 106 structural modules and 52 mechanical modules. Structural modules include all penetrations for piping, cable trays, HVAC duct runs, and all reinforcement for pipe, equipment hangers, and supports. Structural modules are shipped in sub-modules to support transportation by rail or truck or barge. Mechanical modules contain equipment such as pumps, tanks, heat exchangers, air-handling units, and filters along with interconnecting pipes, valves, instruments, wiring and support services. Modular construction requires strong coordination between engineering, supply chain and construction. A total of 8 AP1000 units are currently under construction in China and in the United States. The lessons learned and best practices of each new AP1000 construction are systematically incorporated into the standard design. (A.C.)

25 CFR 1000.161 - What is a self-governance compact?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What is a self-governance compact? 1000.161 Section 1000... EDUCATION ACT Negotiation Process for Annual Funding Agreements Negotiating A Self-Governance Compact § 1000.161 What is a self-governance compact? A self-governance compact is an executed document that affirms...

25 CFR 1000.13 - What is a “nonsignatory Tribe”?

Science.gov (United States)

2010-04-01

... Selection of Additional Tribes for Participation in Tribal Self-Governance Purpose and Definitions § 1000.13... participate in self-governance on its behalf. (2) Meets the eligibility criteria in § 1000.16 but chooses to...) May only become a “signatory Tribe” if it independently meets the eligibility criteria in § 1000.16...

7 CFR 1000.44 - Classification of producer milk.

Science.gov (United States)

2010-01-01

... the market administrator shall determine for each handler described in § 1000.9(a) for each pool plant... the classification of producer milk by allocating the handler's receipts of skim milk and butterfat to the handler's gross utilization of such receipts pursuant to § 1000.43(b)(3) as follows: (a) Skim milk...

Linear and Nonlinear Infrasound Propagation to 1000 km

Science.gov (United States)

2015-12-15

AFRL-RV-PS- AFRL-RV-PS- TR-2016-0017 TR-2016-0017 LINEAR AND NONLINEAR INFRASOUND PROPAGATION TO 1000 KM Catherine de Groot-Hedlin Scripps...Nonlinear Infrasound Propagation to 1000 km 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 62601F 6. AUTHOR(S) Catherine de Groot

SWR 1000 approaching bidding phase status

International Nuclear Information System (INIS)

Brettschuh, W.; Meseth, J.

2000-01-01

The SWR 1000 is an advanced version of the previous 69 and 72 lines by Siemens/KWU. Like its predecesor, the SWR 600, this development project is handled by Siemens on behalf of German utilities, and supported by various European partners contributing engineering and experimental work. With its electric power of approx. 1000 MW, the SWR 1000 differs from earlier KWU boiling water reactor designs especially in having several passive safety systems in addition to some of the active systems customarily used in earlier plants. In a postulated failure of the active safety systems, the passive safety systems can completely take over their function. The passive safety systems are characterized by not needing any external power supply after having been activated, and by not requiring initiation by the I and C system to be activated. In addition, plant operating systems were streamlined, always subject to the guiding principle to employ, as far as possible, technical components and systems with proven operating records. In late 1999, the basic design phase in the development of the SWR 1000 was completed in which all newly designed components and systems as well as those modified greatly over existing BWR plants, including electrical and I and C systems, were designed and the main buildings, such as the reactor building, turbine hall, auxiliary and utilities building, were planned. All newly designed components, systems, and processes were successfully tested for their functioning capability and capacity in large scale tests. The results of the basic design phase were compiled in a site independent safety report based on the rules and regulations applicable in Germany. (orig.) [de

Enhancing AP1000 reactor accident management capabilities for long term accidents

International Nuclear Information System (INIS)

Jiang Pingting; Liu Mengying; Duan Chengjie; Liao Yehong

2015-01-01

Passive safety actions are considered as main measures under severe accident in AP1000 power plant. However, risk is still existed. According to PSA, several probable scenarios for AP1000 nuclear power plant are analyzed in this paper with MAAP the severe accident analysis code. According to the analysis results, several deficiencies of AP1000 severe accident management are found. The long term cooling and containment depressurization capability for AP1000 power plant appear to be most important factors under such accidents. Then, several temporary strategies for AP1000 power plant are suggested, including PCCWST temporary water supply strategy after 72h, temporary injection strategy for IRWST, hydrogen relief action in fuel building, which would improve the safety of AP1000 power plant. At last, assessments of effectiveness for these strategies are performed, and the results are compared with analysis without these strategies. The comparisons showed that correct actions of these strategies would effectively prevent the accident process of AP1000 power plant. (author)

Asymmetric flow events in a VEER 1000

International Nuclear Information System (INIS)

Horak, W.C.; Kennett, R.J.; Shier, W.; Guppy, J.G.

1992-07-01

This paper describes the simulation of asymmetric loss of flow events in Russian designed VVER-1000 reactors using the RETRAN-02 Mod4 computer code. VVER-1000 reactors have significant differences from United States pressurized water reactors including multi-level emergency response systems and plant operation at reduced power levels with one or more main circulation pumps inoperable. The results of these simulations are compared to similar analyses done by the designers for the Rovno plant

Thermohydraulic model of WWER-1000 core

International Nuclear Information System (INIS)

Maroti, L.; Szabados, L.

1987-11-01

Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs

Westinghouse plans global new builds for AP1000

International Nuclear Information System (INIS)

Mitev, Lubomir

2014-01-01

Interview with Danny Roderick, Westinghouse Electric Company, President and Chief Executive Officer since September 2012, about perspectives and future plans for AP1000 new build worldwide. Within three to four years there wille be 'shovels in the ground' for three new AP1000 reactors in the UK, as well as new units in China and Bulgaria. Four AP1000 reactors are under construction in the United States at Vogtle and VC Summer, and soon at Turkey Point. Additionally Danny Roderick spoke about the acquisition of NuGen, technology transfer, the influence of the Ukraine crises on the nuclear market in East Europe and the future need for more nuclear worldwide and in the UK and Bulgaria.

Westinghouse plans global new builds for AP1000

Energy Technology Data Exchange (ETDEWEB)

Mitev, Lubomir [NucNet, Brussels (Belgium)

2014-10-15

Interview with Danny Roderick, Westinghouse Electric Company, President and Chief Executive Officer since September 2012, about perspectives and future plans for AP1000 new build worldwide. Within three to four years there wille be 'shovels in the ground' for three new AP1000 reactors in the UK, as well as new units in China and Bulgaria. Four AP1000 reactors are under construction in the United States at Vogtle and VC Summer, and soon at Turkey Point. Additionally Danny Roderick spoke about the acquisition of NuGen, technology transfer, the influence of the Ukraine crises on the nuclear market in East Europe and the future need for more nuclear worldwide and in the UK and Bulgaria.

ACR-1000TM Project - Licensing Opportunities and Challenges

International Nuclear Information System (INIS)

Popov, N.; Doerffer, S.; Ion, R.; Hopwood, J.

2011-01-01

Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU Reactor TM 1 1000 (ACR-1000 TM ) as an evolutionary advancement of the current CANDU 6 reactor. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU TM systems, components, and materials, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. To ensure that the ACR design is compliant with Canadian and international requirements, regulatory pre-project reviews of the ACR-1000 (and ACR-700 TM 1 with lower output) were conducted early in the design work. The regulatory feedback from these pre-project regulatory reviews helped AECL to better understand regulatory expectations in Canada, US and the UK, and to make further advancements and improvements in the ACR design to meet the Canadian and international regulatory requirements. This paper provides an overview of the key design features of the ACR-1000 reactor design, and summary of the pre-project reviews by those above-mentioned regulatory bodies, demonstrating opportunities and challenges in licensing process of and pointing to the importance of efficient vendor-regulator interaction. (author)

The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

International Nuclear Information System (INIS)

Schene, R.

2009-01-01

Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

L1000FWD: Fireworks visualization of drug-induced transcriptomic signatures.

Science.gov (United States)

Wang, Zichen; Lachmann, Alexander; Keenan, Alexandra B; Ma'ayan, Avi

2018-02-06

As part of the NIH Library of Integrated Network-based Cellular Signatures (LINCS) program, hundreds of thousands of transcriptomic signatures were generated with the L1000 technology, profiling the response of human cell lines to over 20,000 small molecule compounds. This effort is a promising approach toward revealing the mechanisms-of-action (MOA) for marketed drugs and other less studied potential therapeutic compounds. L1000 fireworks display (L1000FWD) is a web application that provides interactive visualization of over 16,000 drug and small-molecule induced gene expression signatures. L1000FWD enables coloring of signatures by different attributes such as cell type, time point, concentration, as well as drug attributes such as MOA and clinical phase. Signature similarity search is implemented to enable the search for mimicking or opposing signatures given as input of up and down gene sets. Each point on the L1000FWD interactive map is linked to a signature landing page, which provides multifaceted knowledge from various sources about the signature and the drug. Notably such information includes most frequent diagnoses, co-prescribed drugs and age distribution of prescriptions as extracted from the Mount Sinai Health System electronic medical records (EMR). Overall, L1000FWD serves as a platform for identifying functions for novel small molecules using unsupervised clustering, as well as for exploring drug MOA. L1000FWD is freely accessible at: http://amp.pharm.mssm.edu/L1000FWD. avi.maayan@mssm.edu. Supplementary data are available at Bioinformatics online. © The Author (2018). Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com

Za koncept vseživljenjskega izobraževanja in vseživljenjskega učenja je koncept izobraževanja odraslih vitalnega pomena

Directory of Open Access Journals (Sweden)

Dušana Findeisen

2012-03-01

Full Text Available Sredi finančne krize se vse bolj izgublja spomin na to, kaj nam je Evropska unija dala doslej in nam daje zdaj, tudi na področju izobraževanja odraslih, zato je morebiti prav, da se vrnemo k preglednim vtisom s konference, ki je zaznamovala deseti rojstni dan programa Grundtvig. Konference Deset let programa Grundtvig ali deset let inovativnosti in inovacij v izobraževanju odrasli v Bruslju sem se resda udeležila kot predstavnica AGE Platform Europe, vodilne evropske organizacije za vprašanja starejših, kjer delujem kot izvedenka za zaposlovanje in izobraževanje starejših. Uradno sicer ne, v resnici pa sem na tej konferenci zagovarjala tudi stališča slovenske univerze za tretje življenjsko obdobje in seveda stališča, ki smo jih razvili v slovenski andragogiki.

24 CFR 1000.104 - What families are eligible for affordable housing activities?

Science.gov (United States)

2010-04-01

... affordable housing activities? 1000.104 Section 1000.104 Housing and Urban Development Regulations Relating... Activities § 1000.104 What families are eligible for affordable housing activities? The following families... Indian area. (b) A non-low income Indian family may receive housing assistance in accordance with § 1000...

Assessment of in-vessel corium retention in CPR1000

International Nuclear Information System (INIS)

Chen Xing; Zhang Shishun; Lin Jiming

2011-01-01

The In-Vessel corium Retention (IVR) strategy of Chinese 1000 MW class commercial pressurized water reactor (CPR1000) is assessed by Risk-Oriented Accident Analysis Methodology (ROAAM). Four representative severe accident scenarios are selected for the IVR assessment in this paper. According to four representative severe accident scenarios consequence calculated by the deterministic code combined with engineering judgment, the input probability distribution of the assessment is determined. Success probability of IVR from the viewpoint of thermal failure is then predicted using MOPOL code. MOPOL is a code developed basing on the well known ROAAM frame and heat transfer model of corium. It is demonstrated that the success probability of IVR by Reactor Cavity Flooding in CPR1000 is potentially higher than 99%. Application of IVR strategy in CPR1000 is envisioned probable if a further more comprehensive risk-benefit evaluation conclusion is positive. (authors)

10 CFR 603.1000 - Contracting officer's responsibilities at time of award.

Science.gov (United States)

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Contracting officer's responsibilities at time of award. 603.1000 Section 603.1000 Energy DEPARTMENT OF ENERGY (CONTINUED) ASSISTANCE REGULATIONS TECHNOLOGY INVESTMENT AGREEMENTS Executing the Award § 603.1000 Contracting officer's responsibilities at time of award...

13 CFR 107.1000 - Licensees without Leverage-exceptions to the regulations.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Licensees without Leverage-exceptions to the regulations. 107.1000 Section 107.1000 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION SMALL BUSINESS INVESTMENT COMPANIES Non-leveraged Licensees-Exceptions to Regulations § 107.1000...

25 CFR 1000.220 - What regulations apply to self-governance Tribes?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What regulations apply to self-governance Tribes? 1000.220 Section 1000.220 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE...-DETERMINATION AND EDUCATION ACT Waiver of Regulations § 1000.220 What regulations apply to self-governance...

AP1000{sup R} licensing and deployment in the United States

Energy Technology Data Exchange (ETDEWEB)

Jordan, R. P.; Russ, P. A.; Filiak, P. P.; Castiglione, L. L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

In recent years, both domestic and foreign utilities have turned to the standardized Westinghouse AP1000 plant design in satisfying their near - and long-term - sustainable energy needs. As direct support to these actions, licensing the AP1000 design has played a significant role by providing one of the fundamental bases in clearing regulatory hurdles leading to the start of new plant construction. Within the U.S. alone, Westinghouse AP1000 licensing activities have reached unprecedented milestones with the approvals of both AP1000 Design Certification and Southern Company's combined construction permit and operating license (COL) application directly supporting the construction of two new nuclear plants in Georgia. Further COL application approvals are immediately pending for an additional two AP1000 plants in South Carolina. And, across the U.S. nuclear industry spectrum, there are 10 other COL applications under regulatory review representing some 16 new plants at 10 sites. In total, these actions represent the first wave of new plant licensing under the regulatory approval process since 1978. Fundamental to the Nuclear Regulatory Commission's AP1000 Design Certification is the formal recognition of the AP1000 passive safety design through regulatory acceptance rulemaking. Through recognition and deployment of the AP1000 Design Certification, the utility licensee / operator of this reactor design are now offered an opportunity to use a simplified 'one-step' combined license process, thereby managing substantial back-end construction schedule risk from regulatory and intervention delays. Application of this regulatory philosophy represents both acceptance and encouragement of standardized reactor designs like the AP1000. With the recent AP1000 Design Certification and utility COL acceptances, the fundamental licensing processes of this philosophy have successfully proven the attainment of significant milestones with the next stage licensing

The 1000 bull genome project

Science.gov (United States)

To meet growing global demands for high value protein from milk and meat, rates of genetic gain in domestic cattle must be accelerated. At the same time, animal health and welfare must be considered. The 1000 bull genomes project supports these goals by providing annotated sequence variants and ge...

24 CFR 1000.8 - May provisions of these regulations be waived?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false May provisions of these regulations be waived? 1000.8 Section 1000.8 Housing and Urban Development Regulations Relating to Housing and... HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.8 May provisions of these...

16 CFR 1000.22 - Office of Human Resources Management.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Office of Human Resources Management. 1000... ORGANIZATION AND FUNCTIONS § 1000.22 Office of Human Resources Management. The Office of Human Resources Management, which is managed by the Director of the Office, provides human resources management support to...

Informational system to assist decision making at spent nuclear fuel transportation from VVER-440, VVER-1000 and RBMK-1000 nuclear power plants

International Nuclear Information System (INIS)

Kuryndin, A.V.; Kirkin, A.M.; Stroganov, A.A.

2012-01-01

The developed informational system provides an automated estimations of nuclear and radiation safety parameters during spent nuclear fuel transportation from WWER-440 and WWER-1000 and RBMK-1000 nuclear power plants to the nuclear fuel cycle facilities, and allows us to determine the optimum cask loading from the dose rates distribution outside of protection point of view [ru

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

1995-12-31

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.)

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Ubra, O [Skoda Company, Prague (Switzerland); Doubek, M [Czech Technical Univ., Prague (Switzerland)

1996-12-31

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

Vver-1000 Mox core computational benchmark

International Nuclear Information System (INIS)

2006-01-01

The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

7 CFR 1000.4 - Plant.

Science.gov (United States)

2010-01-01

... transferring bulk milk from one tank truck to another or a separate building used only as a distribution point... § 1000.4 Plant. (a) Except as provided in paragraph (b) of this section, plant means the land, buildings, facilities, and equipment constituting a single operating unit or establishment at which milk or milk...

Influence of High Pressure and High Temperature Hydrogen on Fracture Toughness of Ni-Containing Steels and Alloys

Directory of Open Access Journals (Sweden)

Balitskii Alexander

2014-03-01

Full Text Available W ramach przeprowadzonych badan okreslono wpływ wodoru na krótkotrwała wytrzymałosc i płasko-naprezeniowa ciagliwosc stali 10Cr15Ni27 oraz stopów niklu 04Cr16Ni56, 05Cr19Ni55 przy wielkosciach cisnienia do 35 MPa i zakresie temperatury od 293 do 773 K. Wpływ wodoru wysokiego cisnienia powoduje, ze wydłuzenie, plastycznosc, wytrzymałosc nisko cyklowa N i parametr odpornosci na pekanie Kc przyjmuja minimalne wartosci dla próbek stopu 04Cr16Ni56 przy cisnieniu wynoszacym 10 MPa, natomiast dla nasyconych wodorem próbek stali 10Cr15Ni27 i stopu 05Cr19Ni55 minimum wartosci tych parametrów wystepuje przy cisnieniu 15 MPa, dla koncentracji wodoru wynoszacej odpowiednio 15 i 19 wppm.

Analysis of operating reliability of WWER-1000 unit

International Nuclear Information System (INIS)

Bortlik, J.

1985-01-01

The nuclear power unit was divided into 33 technological units. Input data for reliability analysis were surveys of operating results obtained from the IAEA information system and certain indexes of the reliability of technological equipment determined using the Bayes formula. The missing reliability data for technological equipment were used from the basic variant. The fault tree of the WWER-1000 unit was determined for the peak event defined as the impossibility of reaching 100%, 75% and 50% of rated power. The period was observed of the nuclear power plant operation with reduced output owing to defect and the respective time needed for a repair of the equipment. The calculation of the availability of the WWER-1000 unit was made for different variant situations. Certain indexes of the operating reliability of the WWER-1000 unit which are the result of a detailed reliability analysis are tabulated for selected variants. (E.S.)

Discussion of QA grading for AP1000 NP plant

International Nuclear Information System (INIS)

Luo Shuiyun; Zhang Qingchuan

2012-01-01

The grading method of quality assurance for the following AP1000 project is presented based on the Westinghouse classification principle, referring to the classification method of the AP1000 self-reliance supporting project and considering the factors of classification, which can meet the requirements of domestic nuclear safety regulation and standard of the QA classification. (authors)

Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

2014-01-01

Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

24 CFR 1000.58 - Are there limitations on the investment of IHBG funds?

Science.gov (United States)

2010-04-01

... investment of IHBG funds? 1000.58 Section 1000.58 Housing and Urban Development Regulations Relating to..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.58 Are there... its rights under § 1000.60. (e) Expenditure of funds for affordable housing activities under section...

Westinghouse AP1000® PWR: Meeting Customer Commitments and Market Needs

International Nuclear Information System (INIS)

Shulyak, Nick

2014-01-01

Westinghouse Electric Company once again sets a new industry standard with the AP1000 reactor. Historically, Westinghouse plant designs and technology have forged the cutting edge of worldwide nuclear technology. Today, about 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory Commission (NRC). The AP1000 features proven technology, innovative passive safety systems and offers: Unequalled safety, Economic competitiveness, Improved and more efficient operations. The AP1000 builds and improves upon the established technology of major components used in current Westinghouse-designed plants with proven, reliable operating experience over the past 50 years. These components include: Steam generators, Digital instrumentation and controls, Fuel, Pressurizers, Reactor vessels. Simplification was a major design objective for the AP1000. The simplified plant design includes overall safely systems, normal operating systems, the control room, construction techniques, and instrumentation and control systems. The result is a plant that is easier and less expensive to build, operate and maintain. The AP1000 design saves money and time with an accelerated construction time period of approximately 36 months, from the pouring of first concrete to the loading of fuel. Also, the innovative AP1000 features: 50% fewer safety-related valves, 80% less safety-related piping, 85% less control cable, 35% fewer pumps , 45% less seismic building volume. Eight AP1000 units under construction worldwide-Four units in China-Four units in the United States. (author)

The passive safety systems of the Swr 1000

International Nuclear Information System (INIS)

Neumann, D.

2001-01-01

In recent years, a new boiling water reactor (BWR) plant called the SWR 1000 has been developed by Siemens on behalf of Germany's electric utilities. This new plant design concept incorporates the wide range of operating experience gained with German BWRs. The main objective behind developing the SWR 1000 was to design a plant with a rated electric output of approximately 1000 MW which would not only have a lower capital cost and lower power generating costs but would also provide a much higher level of nuclear safety compared to plants currently in operation. This safety-related goal has been met through, for example, the use of passive safety equipment. Passive systems make a significant contribution towards increasing the over-all level of plant safety due to the way in which they operate. They function solely accord-ing to basic laws of nature, such as gravity, and perform their designated functions with-out any need for electric power or other sources of external energy, or signals from instrumentation and control (I and C) equipment. The passive safety systems have been designed such that design basis accidents can be controlled using just these systems alone. However, the design concept of the SWR 1000 is nevertheless still based on the provision of active safety systems in addition to passive systems. (author)

MDEP AP1000WG Design-Specific Common Position CP-AP1000WG-02. Common position addressing Fukushima Daiichi NPP accident-related issues

International Nuclear Information System (INIS)

2016-09-01

A severe accident involving several units took place in Japan at Fukushima Daiichi nuclear power plant (NPP) in March 2011. The immediate cause of the accident was an earthquake followed by a tsunami coupled with inadequate provisions against the consequences of such events in the design. Opportunities to improve protection against a realistic design basis tsunami had not been taken. As a consequence of the tsunami, safety equipment and the related safety functions were lost at the plant, leading to core damage in three units and subsequently to large radioactive release. Several studies have already been performed to better understand the accident progression and detailed technical studies are still in progress in Japan and elsewhere. In the meantime, on-going studies on the behaviour of nuclear power plants in very severe situations, similar to Fukushima Daiichi, seek to identify potential vulnerabilities in plant design and operation; to suggest reasonably practicable upgrades; or to recommend enhanced regulatory requirements and guidance to address such situations. Likewise, agencies around the world that are responsible for regulating the design, construction and operation of AP1000 R plants are engaged in similar activities. The MDEP AP1000 R Working Group (AP1000 WG) members consist of members from Canada, China, the United Kingdom and the United States. Since the regulatory review of their AP1000 R applications have not been completed by all of these Countries yet, this paper identifies common preliminary approaches to address potential safety improvements for AP1000 R plants as related to lessons learned from the Fukushima Daiichi accident or Fukushima Daiichi-related issues. In seeking common position, regulators will provide input to this paper to reflect their safety conclusions regarding the AP1000 R design and how the design could be enhanced to address Fukushima Daiichi issues. The common preliminary approaches are organized into five sections

29 CFR 1926.1000 - Rollover protective structures (ROPS) for material handling equipment.

Science.gov (United States)

2010-07-01

... 29 Labor 8 2010-07-01 2010-07-01 false Rollover protective structures (ROPS) for material handling equipment. 1926.1000 Section 1926.1000 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY... CONSTRUCTION Rollover Protective Structures; Overhead Protection § 1926.1000 Rollover protective structures...

Safety analysis methodology for OPR 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun

2005-01-01

Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

Study on modular construction management in AP1000 nuclear plant project

International Nuclear Information System (INIS)

Fang Xiaopeng; Shen Wenrong; Sun Kebin; Wei Zhong

2010-01-01

The construction of AP1000 Nuclear Power Plant (NPP) has commenced in China. The AP1000 NPP features a passive design concept and modular construction technology. Based on the management of the construction of current AP1000 NNP, this paper describes the effects on Nuclear Island (NI) construction project management resulting from modular construction technology, as well as new construction techniques and methods. This paper puts forward new requirements for construction schedule management of the nuclear island construction at different levels. The AP1000 NI construction logic features the parallel construction of civil and structural erection as the main approach, with the integrated schedule of module fabrication, assembly and installation as support. The structural modules of AP1000 project are prefabricated in shop, delivered to site as sub-modules and assembled to integrated structural module. The assembled module is transported to the construction site, hoisted and finally set in NI. This paper illustrates how to ensure the construction quality of structural modules by analyzing the interface process and key links in the quality control program, and introduces how to ensure the safety of heavy structural components during various construction phases by evaluating and analyzing the construction safety process. This paper also makes an analysis of the safe environment for the assembly and installation of Containment Vessel, the management of product protection and personnel safety inside the Containment Building during 'Open Top' construction, raises to implement effective protection for the numerous pre-set mechanical modules and equipments, as well as personnel safety protection programs and measures. The modular construction feature of AP1000 NPP design requires technique improvement and management innovation during the NI construction. This paper makes a study and research on the control management of schedule, quality and safety of AP1000 NPP NI

QTL mapping and correlation analysis for 1000-grain weight and ...

Indian Academy of Sciences (India)

in both environments, nine QTL for 1000-paddy-grain weight (PTGW), five QTL for 1000-brown-grain weight .... at the middle of chromosome 4 (defined by Bb38P21a), one ..... tive traits for panicle architecture by using chromosomal segment.

46 CFR 153.1000 - Special operating requirements for cargoes reactive with water.

Science.gov (United States)

2010-10-01

... 46 Shipping 5 2010-10-01 2010-10-01 false Special operating requirements for cargoes reactive with water. 153.1000 Section 153.1000 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED... MATERIALS Operations Special Cargo Procedures § 153.1000 Special operating requirements for cargoes reactive...

Finite element modeling of AP1000 nuclear island

International Nuclear Information System (INIS)

Tinic, S.; Orr, R.

2003-01-01

The AP1000 is a standard design developed by Westinghouse and its partners for an advanced nuclear power plant utilizing passive safety features. It is based on the certified design of the AP600 and has been uprated to 1000 MWe. The plant has five principal building structures; the nuclear island, the turbine building; the annex building; the diesel generator building and the radwaste building. The nuclear island consists of the containment building (the steel containment vessel and the containment internal structures), the shield building, and the auxiliary building. These structures are founded on a common basemat and are collectively known as the nuclear island. This paper describes use of the general purpose finite element program ANSYS [2] in structural analyses and qualification of the AP1000 nuclear island buildings. It describes the modeling of the shield building and the auxiliary building and the series of analyses and the flow of information from the global analyses to the detailed analyses and building qualification. (author)

Natures and incidences of each normal variations on I.V.P. series of 1000 Korean females

International Nuclear Information System (INIS)

Shin, Ok Ja; Chin, Soo Yil

1980-01-01

Each organ of body has many normal variation and/or abnormal developments from fetus to adult, especially urogenital system is very complex and markedly abnormal. The authors attempt to analyze the natures, incidences and differences of each normal variations based on I.V.P. series of 1000 Korean females and the review of literatures were carried out. The results are as follows: 1. The normal variations of urinary tracts are about 30 kinds as follows; Aphasia 0/1000 (0%), Hypoplasia 4/1000 (0.4%), Atrophic kidney 0/1000 (0%), Hyperplasia 0/1000 (0%), Hypertrophied kidney 0/1000 (0%), Supernumerary kidney 0/1000 (0%), Pelvic kidney 0/1000 (0%), Wandering kidney 7/1000 (7%), Horseshoe kidney 2/1000 (2%), Nephroptosis 6/1000 (0.6%), Malroated kidney 0/1000 (0%), Crossed ecotopic kidney 0/1000 (0%), Ureteral kingking 440/1000 (44.0%), Short ureter 3/1000 (0.3%), Long ureter 49/1000 (4.9%), Retrocaval ureter 0/1000 (0%), Retroiliac ureter 0/1000 (0%), Mega-ureter 0/1000 (0%), Aberrant vessel in ureter 79/1000 (7.9%), Congenital ureteral valves 0/1000 (0%), Anomalous ureteral orifices 0/1000 (0%), Double ureter 22/1000 (2.2%), Ureteral diverticulum 0/1000 (0%), Extrarenal pelvis 170/1000 (17.0%), Intrarenal pelvis 830/1000 (83.0%), Supernumerary pelvis 55/1000 (55%), Renal backflow 2/1000 0.2%), Multiple calyces 0/1000 (0.1%), Calyceal diverticulum 2/1000 (0.2%), Unilateral fused kidney 0/1000 (0%), Polar enlarged kidney 0/1000 (0%). 2. The order of incidence of common variations in Korean females are as follows: 1. Intrarenal pelvis 2. Ureteral kingking 3. Extrarenal pelvis 4. Aberrant vessel in ureter 5. Double pelvis 6. Long ureter 7. Double ureter 8. Wandering Kidney 9. Nephrotosis 10. Hypoplasia 11. Short ureter 12. Horseshoe kidney 13. Renal backflow 14. Calyceal diverticulum 15. Supernumerary calyces. 3. The most common variations in Korean females are intrarenal pelvis, extrarenal pelvis, aberrant vessel in ureters and most of them are acquired origin. Contrary

Comparison Study of Water Demineralization System for the OPR 1000 and AP 1000 Nuclear Power Plant

International Nuclear Information System (INIS)

Dedy Priambodo; Siti Alimah; Erlan Dewita

2009-01-01

OPR 1000 adopts demineralization method based on ion exchanger resin and AP 1000 adopt the method that based on Reverse Osmosis (RO)-Electrodeionization (EDI). The Ion exchange process is a reversible chemical reaction of a solution and an insoluble solid. Ion exchanger use resin as polluter ions capture and will be regenerated after its saturated. RO is method using pressure to force a solution through a membrane, retaining the solute on one side and allowing the pure solvent to pass to the other side. Whereas, EDI is a combination of ion exchange and electrodialysis. The ions is taken by ion exchange resin, and then it is discharged utilizing electric potential difference. Due to water splitting phenomena in EDI, make resin will never be saturated, so the RO-EDI process is water demineralization system that use little chemical, more simple installation, capable to maintain demineralization water product quality and environmental friendly. Thereby, The RO-EDI water demineralization system is more advance then ion exchange technology. (author)

1000 meters water depth rigid TLP riser; Riser rigido de plataforma de pernas atirantadas para lamina d'agua de 1000 metros

Energy Technology Data Exchange (ETDEWEB)

Braga, Mauro Jacinto Pastor

1990-07-01

A procedure to estimate the fatigue life of a TLP riser in 1000 meters water depth based on a hydro-elastic analysis of an integrated riser-TLP model in the time domain is presented . The computational architecture is shown that makes it feasible to process and store the great amount of data involved. The procedure is applied to a 1000 meters water depth TLP with a set of 40 risers 8 inches in diameter equipped with a floatation layer. (author)

16 CFR 1000.21 - Office of Compliance and Field Operations.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Office of Compliance and Field Operations. 1000.21 Section 1000.21 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION GENERAL COMMISSION... addressed through rulemaking or voluntary standards. The Office develops surveillance strategies and...

24 CFR 1000.30 - What prohibitions regarding conflict of interest are applicable?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What prohibitions regarding conflict of interest are applicable? 1000.30 Section 1000.30 Housing and Urban Development Regulations... HOUSING, DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.30...

Performance of the Westinghouse WWER-1000 fuel design

International Nuclear Information System (INIS)

Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

2011-01-01

In 2005 six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in the South Ukraine Unit 3. This design has demonstrated full compatibility with resident fuel designs and all associated fuel handling and reactor components. Operations have further demonstrated adequacy of performance margins and the reliability requirements for multiple cycles of operation. The LTA's have now been discharged after completing the planned four cycles of operation and having reached an average assembly burnup in excess of 43 MWd/kgU. Post Irradiation Examinations were performed after completion of each cycle. The final LTA inspection program at end of Cycle 20 in 2010 yielded satisfactory results on all counts, and it was concluded that the 6 Westinghouse LTA's performed as expected during their operational regimes. Very good performance was demonstrated in the WWER-1000 reactor environment for the Zr-1%Nb as grid material, and ZIRLO fuel cladding and structural components. Control Rod Assemblies drop times and drag forces were all within the accepted values. The LTA program demonstrated that this fuel design is suitable for full core applications. However, the topic of fuel assembly distortion resistance was re-visited and Westinghouse therefore considered operational experience and design features from multiple development programs to enhance the basic Westinghouse WWER-1000 fuel design for Ukrainian reactors. The design now includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. This paper describes briefly the development of the Westinghouse WWER-1000 fuel design and how test results and operational experiences from multiple sources have been utilized to produce a most suitable fuel design. Early in 2011 a full region of the Westinghouse WWER-1000 design completed another full cycle of operation at South Ukraine Unit 3, all with excellent results. All 42 fuel assemblies were examined for visible damage or non

Complications after 1000 lung radiofrequency ablation sessions in 420 patients: a single center's experiences.

Science.gov (United States)

Kashima, Masataka; Yamakado, Koichiro; Takaki, Haruyuki; Kodama, Hiroshi; Yamada, Tomomi; Uraki, Junji; Nakatsuka, Atsuhiro

2011-10-01

This study retrospectively evaluates complications after lung radiofrequency ablation (RFA). Complications were assessed for each RFA session in 420 consecutive patients with 1403 lung tumors who underwent 1000 RFA sessions with a cool-tip RFA system. A major complication was defined as a grade 3 or 4 adverse event. Risk factors affecting frequent major complications that occurred with an incidence of 1% or more were detected using multivariate analysis. Four deaths (0.4% [4/1000]) related to RFA procedures occurred. Three patients died of interstitial pneumonia. The other patient died of hemothorax. The major complication rate was 9.8% (98/1000). Frequent major complications were aseptic pleuritis (2.3% [23/1000]), pneumonia (1.8% [18/1000]), lung abscess (1.6% [16/1000]), bleeding requiring blood transfusion (1.6% [16/1000]), pneumothorax requiring pleural sclerosis (1.6% [16/1000]), followed by bronchopleural fistula (0.4% [4/1000]), brachial nerve injury (0.3% [3/1000]), tumor seeding (0.1% [1/1000]), and diaphragm injury (0.1% [1/1000]). Puncture number (p risk factors for aseptic pleuritis. Previous external beam radiotherapy (p risk factors for pneumonia, as were emphysema (p lung abscess, and serum platelet count (p risk factor for pneumothorax requiring pleural sclerosis. Lung RFA is a relatively safe procedure, but it can be fatal. Risk factors found in this study will help to stratify high-risk patients.

HLA diversity in the 1000 genomes dataset.

Directory of Open Access Journals (Sweden)

Pierre-Antoine Gourraud

Full Text Available The 1000 Genomes Project aims to provide a deep characterization of human genome sequence variation by sequencing at a level that should allow the genome-wide detection of most variants with frequencies as low as 1%. However, in the major histocompatibility complex (MHC, only the top 10 most frequent haplotypes are in the 1% frequency range whereas thousands of haplotypes are present at lower frequencies. Given the limitation of both the coverage and the read length of the sequences generated by the 1000 Genomes Project, the highly variable positions that define HLA alleles may be difficult to identify. We used classical Sanger sequencing techniques to type the HLA-A, HLA-B, HLA-C, HLA-DRB1 and HLA-DQB1 genes in the available 1000 Genomes samples and combined the results with the 103,310 variants in the MHC region genotyped by the 1000 Genomes Project. Using pairwise identity-by-descent distances between individuals and principal component analysis, we established the relationship between ancestry and genetic diversity in the MHC region. As expected, both the MHC variants and the HLA phenotype can identify the major ancestry lineage, informed mainly by the most frequent HLA haplotypes. To some extent, regions of the genome with similar genetic or similar recombination rate have similar properties. An MHC-centric analysis underlines departures between the ancestral background of the MHC and the genome-wide picture. Our analysis of linkage disequilibrium (LD decay in these samples suggests that overestimation of pairwise LD occurs due to a limited sampling of the MHC diversity. This collection of HLA-specific MHC variants, available on the dbMHC portal, is a valuable resource for future analyses of the role of MHC in population and disease studies.

AP1000 Containment Design and Safety Assessment

International Nuclear Information System (INIS)

Wright, Richard F.; Ofstun, Richard P.; Bachere, Sebastien

2002-01-01

The AP1000 is an up-rated version of the AP600 passive plant design that recently received final design certification from the US NRC. Like AP600, the AP1000 is a two-loop, pressurized water reactor featuring passive core cooling and passive containment safety systems. One key safety feature of the AP1000 is the passive containment cooling system which maintains containment integrity in the event of a design basis accident. This system utilizes a high strength, steel containment vessel inside a concrete shield building. In the event of a pipe break inside containment, a high pressure signal actuates valves which allow water to drain from a storage tank atop the shield building. Water is applied to the top of the containment shell, and evaporates, thereby removing heat. An air flow path is formed between the shield building and the containment to aid in the evaporation and is exhausted through a chimney at the top of the shield building. Extensive testing and analysis of this system was performed as part of the AP600 design certification process. The AP1000 containment has been designed to provide increased safety margin despite the increased reactor power. The containment volume was increased to accommodate the larger steam generators, and to provide increased margin for containment pressure response to design basis events. The containment design pressure was increased from AP600 by increasing the shell thickness and by utilizing high strength steel. The passive containment cooling system water capacity has been increased and the water application rate has been scaled to the higher decay heat level. The net result is higher margins to the containment design pressure limit than were calculated for AP600 for all design basis events. (authors)

1,000 Wells for Darfur

Science.gov (United States)

Virji, Hassan

2007-08-01

A new humanitarian mission called ``1,000 Wells for Darfur'' grew out of the discovery from recent space data of an ancient megalake in a large basin in that region. Eman Ghoneim, a research professor at Boston University's Center for Remote Sensing, and Center director Farouk El-Baz mapped the ancient lake's boundary using Landsat, RADARSAT, and Shuttle Radar Topography Mission (SRTM) data.

21 CFR 1000.60 - Recommendation on administratively required dental x-ray examinations.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Recommendation on administratively required dental x-ray examinations. 1000.60 Section 1000.60 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT... Recommendations § 1000.60 Recommendation on administratively required dental x-ray examinations. (a) The Food and...

24 CFR 1000.48 - Are Indian preference requirements applicable to IHBG activities?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Are Indian preference requirements applicable to IHBG activities? 1000.48 Section 1000.48 Housing and Urban Development Regulations Relating to..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.48 Are Indian...

24 CFR 1000.128 - Is income verification required for assistance under NAHASDA?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Is income verification required for assistance under NAHASDA? 1000.128 Section 1000.128 Housing and Urban Development Regulations Relating to... § 1000.128 Is income verification required for assistance under NAHASDA? (a) Yes, the recipient must...

24 CFR 1000.52 - What Indian preference requirements apply to IHBG procurement?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What Indian preference requirements apply to IHBG procurement? 1000.52 Section 1000.52 Housing and Urban Development Regulations Relating to..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.52 What Indian...

The ELIXIR channel in F1000Research.

Science.gov (United States)

Blomberg, Niklas; Oliveira, Arlindo; Mons, Barend; Persson, Bengt; Jonassen, Inge

2015-01-01

ELIXIR, the European life science infrastructure for biological information, is a unique initiative to consolidate Europe's national centres, services, and core bioinformatics resources into a single, coordinated infrastructure. ELIXIR brings together Europe's major life-science data archives and connects these with national bioinformatics infrastructures  - the ELIXIR Nodes. This editorial introduces the ELIXIR channel in F1000Research; the aim of the channel is to collect and present ELIXIR's scientific and operational output, engage with the broad life science community and encourage discussion on proposed infrastructure solutions. Submissions will be assessed by the ELIXIR channel Advisory Board to ensure they are relevant to ELIXIR community, and subjected to F1000Research open peer review process.

Advanced fuel cycles for WWER-1000 reactors

International Nuclear Information System (INIS)

Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

2007-01-01

Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

Fuel improvement and WWER-1000 FA main operational results

International Nuclear Information System (INIS)

Rozhkov, V.; Enin, A.; Bezborodov, Y.; Petrov, V.

2003-01-01

The JSC NCCP experience of WWER-1000 Fuel Assemblies (FAs) fabrication and operation confirms the adequate feasibility and efficiency of fuel operation in 3-4-x fuel cycles, high operating reliability and competitive capacity as compared with foreign analogues. The work on fuel improvement is aimed at an improvement of the operating reliability and an enhancement of the fuel use efficiency in WWER-1000 advanced FAs

AP1000: Meeting economic goals in a competitive world. Annex 7

International Nuclear Information System (INIS)

Davis, G.; Cummins, E.; Winters, J.

2002-01-01

In the U.S., conditions are becoming more favorable for considering the nuclear option again for new baseload generation. While oil and natural gas prices have risen, the cost of operating the existing fleet of nuclear plants has decreased. Furthermore, an advanced 1000 MWe nuclear plant that will be even more cost-competitive with fossil fuels and natural gas will be available by 2005. Westinghouse, in an effort to further improve on the AP600's cost competitiveness, has developed the AP1000, a two-loop, 1000 MWe, advanced pressurized water reactor (PWR) with passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. Like the AP600, the AP1000 uses proven technology that builds on over 30 years of operating PWR experience. Westinghouse has completed design studies that demonstrate that it is feasible to increase the power output of the AP600 to at least 1000 MWe, maintaining its current design configuration and licensing basis. To maximize the cost savings, the AP1000 has been designed within the space constraints of the AP600, while retaining the credibility of proven components and substantial safety margins. The affect on the plant's overnight cost of the increased major components that is required to uprate the AP600 to 1000 MWe is small. This overall cost addition is on the order of 11 percent, while the overall power increase is almost 80 percent. This paper describes the changes made to uprate the AP600 and gives an overview of the plant design. (author)

DCS emulator development for the ACR-1000 simulator

International Nuclear Information System (INIS)

Nakashima, Y.; Trueman, R.; Ishii, K.

2010-01-01

Nuclear Power Plant (NPP) simulators are the main means for operator training and as such are a crucial part of the NPP operation life-cycle. Hitachi, Ltd., Information and Control Systems Company (henceforth 'Hitachi') is the preferred DCS and DCS emulator developer and supplier for the ACR-1000 NPP control system. Hitachi's concept for the DCS (distributed control system) portion of the ACR-1000 simulator is 'total emulation of the DCS' by software. This paper will review the current status of the technical development and the major project milestones. (author)

24 CFR 1000.6 - What is the nature of the IHBG program?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What is the nature of the IHBG program? 1000.6 Section 1000.6 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.6 What is the nature of the IHBG...

25 CFR 115.1000 - Who owns the records associated with this part?

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Who owns the records associated with this part? 115.1000 Section 115.1000 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR FINANCIAL ACTIVITIES TRUST FUNDS FOR TRIBES AND INDIVIDUAL INDIANS Records § 115.1000 Who owns the records associated with this...

Probability Analysis of the Construction Cost of an APR1000 Single Unit

International Nuclear Information System (INIS)

Ha, Gak Hyeon; Suh, Yong Pyo; Kang, Yong Chul

2011-01-01

The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, cutting CO 2 emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the APR1000 development is to ensure export competitiveness in the developing countries in the Middle East and Southeast Asia. To that end, APR1000 (1,000MWe, 3 rd generation) will be developed based on the OPR1000 (Korean standard nuclear power plant, 2.5 generation) by incorporating and improving the general requirements such as the 60 year design life time, comprehensive site requirement of 0.3g seismic design, stability improvement, operability improvement and provisions for severe accidents. The APR1000 adds 16 advanced design features to its predecessor, as outlined below in Table 1

25 CFR 166.1000 - Who owns the records associated with this part?

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Who owns the records associated with this part? 166.1000 Section 166.1000 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR LAND AND WATER GRAZING PERMITS Records § 166.1000 Who owns the records associated with this part? (a) Records are the property of...

16 CFR 1000.12 - Organizational structure.

Science.gov (United States)

2010-01-01

... FUNCTIONS § 1000.12 Organizational structure. The Consumer Product Safety Commission is composed of the... Commission: (1) Office of the General Counsel; (2) Office of Congressional Relations; (3) Office of the... Reduction; (3) Office of Information and Public Affairs; (4) Office of Compliance and Field Operations; (5...

ASD-1000: High-resolution, high-temperature acetylene spectroscopic databank

Science.gov (United States)

Lyulin, O. M.; Perevalov, V. I.

2017-11-01

We present a high-resolution, high-temperature version of the Acetylene Spectroscopic Databank called ASD-1000. The databank contains the line parameters (position, intensity, Einstein coefficient for spontaneous emission, term value of the lower states, self- and air-broadening coefficients, temperature dependence exponents of the self- and air-broadening coefficients) of the principal isotopologue of C2H2. The reference temperature for line intensity is 296 K and the intensity cutoff is 10-27 cm-1/(molecule cm-2) at 1000 K. The databank has 33,890,981 entries and covers the 3-10,000 cm-1 spectral range. The databank is based on the global modeling of the line positions and intensities performed within the framework of the method of effective operators. The parameters of the effective Hamiltonian and the effective dipole moment operator have been fitted to the observed values of the line positions and intensities collected from the literature. The broadening coefficients as well as their temperature dependence exponents were calculated using the empirical equations. The databank is useful for studying high-temperature radiative properties of C2H2. ASD-1000 is freely accessible via the Internet site of V.E. Zuev Institute of Atmospheric Optics SB RAS ftp://ftp.iao.ru/pub/ASD1000/.

HSE management for AP1000 nuclear plant construction in EPC mode

International Nuclear Information System (INIS)

He Xiaogang; Wei Zhong

2010-01-01

As a new nuclear type, AP1000 will become the development direction of Chinese nuclear project. EPC General Contract mode is favored by nuclear owners both at home and abroad. Therefore, there is necessity for studying HSE management system and method suitable for AP1000 nuclear plant construction (ANPC) based on combination of AP1000 construction characters in EPC mode. This can not only ensure safety for ANPC but also positively promote national nuclear power development. For this reason, based on site HSE management of the first AP1000 nuclear plant under construction, HSE management system and method for ANPC in EPC mode was proposed after analysis of the character of EPC mode and ANPC character. It is hoped that it will be helpful for safe construction for ANPC. (authors)

Evaluation of images in a DR system (KD-1000) with new photostimulable phosphors

International Nuclear Information System (INIS)

Yoshita, Hisashi; Kuranishi, Makoto; Nakamura, Mamoru; Kato, Syosuke; Matsui, Mitate.

1994-01-01

The Konica Direct Digitizer KD-1000 is a compact chest acquisition unit in digital radiography. The KD-1000 system uses a photostimulable phosphor detector (RbBr: Tl) and directly converts X-ray quantum into digitized information. We have studied the physical characteristics of the KD-1000, which consists of input-output relationship, spatial resolution and noise. Additionally, we have made a comparison of visual detectability between images from the KD-1000 system and those from a film/screen system (KO-380/HRL), which is used for chest radiography. We have quantified these basic characteristics and established the advantage of the KD-1000 system. (author)

Evaluation of images in a DR system (KD-1000) with new photostimulable phosphors

Energy Technology Data Exchange (ETDEWEB)

Yoshita, Hisashi; Kuranishi, Makoto; Nakamura, Mamoru; Kato, Syosuke (Toyama Medical and Pharmaceutical Univ. (Japan). Hospital); Matsui, Mitate

1994-01-01

The Konica Direct Digitizer KD-1000 is a compact chest acquisition unit in digital radiography. The KD-1000 system uses a photostimulable phosphor detector (RbBr: Tl) and directly converts X-ray quantum into digitized information. We have studied the physical characteristics of the KD-1000, which consists of input-output relationship, spatial resolution and noise. Additionally, we have made a comparison of visual detectability between images from the KD-1000 system and those from a film/screen system (KO-380/HRL), which is used for chest radiography. We have quantified these basic characteristics and established the advantage of the KD-1000 system. (author).

Inner-Outer Lane Advantage in Olympic 1000 Meter Speed Skating

NARCIS (Netherlands)

Kamst, Richard; Kuper, Gerard H.; Sierksma, Gerard; Talsma, Bertus G.

During the Olympic Games and the World Championships Single Distances the 1000m is skated by every skater only one time. However, there may be a difference in skating a 1000m race with a start in the inner and the outer lane that introduces an externality that introduces unfairness. We show that

25 CFR 1000.105 - What are self-governance base budgets?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What are self-governance base budgets? 1000.105 Section... Base Budgets § 1000.105 What are self-governance base budgets? (a) A Tribe/Consortium self-governance base budget is the amount of recurring funding identified in the President's annual budget request to...

Concentration of WWER-1000 unit power on one site

International Nuclear Information System (INIS)

Rousek, J.; Kysel, J.; Sladek, V.

1987-01-01

The problem of a suitable number of nuclear power plant units built on one site is discussed. Using an example of three sites being prepared now in Czechoslovakia, two alternatives - one with two WWER-1000 units, the other with four WWER-1000 units on one site - are evaluated from the viewpoint of long-range nuclear power development program in Czechoslovakia, costs, transmission of electric power and heat supply. (author). 10 tabs., 13 refs

Estonskije filmõ stali laureatami festivalei

Index Scriptorium Estoniae

2008-01-01

Sõltumatu kino 35. festivalil Brüsselis sai peaauhinna Veiko Õunpuu "Sügisball" ja animafilmifestivalil Utrechtis Hollandis võitis peapreemia Priit ja Olga Pärna joonisfilm "Elu ilma Gabriella Ferrita"

48 CFR 3.1000 - Scope of subpart.

Science.gov (United States)

2010-10-01

... BUSINESS PRACTICES AND PERSONAL CONFLICTS OF INTEREST Contractor Code of Business Ethics and Conduct 3.1000... codes of business ethics and conduct, and display of agency Office of Inspector General (OIG) fraud...

Influence of performance improvement of AP1000 nuclear island main equipment forging on manufacturing

International Nuclear Information System (INIS)

Liu Zhiying

2013-01-01

In order to comply with the 60-year design life of an AP1000 nuclear power station, higher strength and ductility requirements have been made on AP1000 nuclear island SG forgings than on CPR1000 nuclear island main equipment. In addition, bigger size of AP1000 nuclear island SG forgings increases the difficulty of manufacturing them. Insufficient recognition of these changes may cause unstable quality of forgings and possible quality problems in follow-up welding procedure. On the basis of comparison and analysis of AP1000 nuclear island SG forgings and CPR1000 nuclear island forgings, this thesis suggests clear directions for the actions we need to take. (author)

24 CFR 1000.32 - May exceptions be made to the conflict of interest provisions?

Science.gov (United States)

2010-04-01

... conflict of interest provisions? 1000.32 Section 1000.32 Housing and Urban Development Regulations Relating... HOUSING, DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.32... conflict of interest provisions set forth in § 1000.30(b) on a case-by-case basis when it determines that...

FIDEDIGNIDADE DA CORRIDA DE 1000 METROS PARA ESPAÇO REDUZIDO

Directory of Open Access Journals (Sweden)

Marilia do Rosso Krug

2008-07-01

Full Text Available Este estudo teve como objetivo estabelecer a fidedignidade do teste de corrida de 1000 metros, padronizado por KLISSOURAS (1973, para espaço reduzido. A amostra foi constituída aleatoriamente de 240 escolares, sendo 120 escolares de cada sexo, e 30 escolares por idade e sexo, nas faixas etárias de 9 a 12 anos, matriculados nas escolas da rede municipal de ensino, da zona urbana, da cidade de Pelotas, RS. O peso corporal foi medido através de uma balança de alavanca e a estatura pelo estadiômetro. Utilizou-se o teste de corrida de 1000 metros, padronizado por KLISSOURAS (1973, para medir a potência aeróbica. Obteve-se correlação positiva significativa entre os resultados do teste de corrida de 1000 metros realizados em pista de atletismo com o de espaço reduzido (configuração oval. Conclui-se que na falta da pista de atletismo, pode-se realizar o teste de corrida de 1000 metros em espaço reduzido (configuração oval.

Burnup dependence of coolant void reactivity for ACR-1000 cell

International Nuclear Information System (INIS)

Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

2008-01-01

The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

SWR 1000: Efficient design for operational excellence

International Nuclear Information System (INIS)

Brettschuh, W.

2008-01-01

The SWR 1000 boiling water reactor (BWR) offers all of the advantages associated with an advanced plant design, i.e. excellent safety performance and competitive power generation costs, in the medium capacity range (1000-1250 MW). The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by using very simple passive safety equipment, simplified plant operating systems, and a very simple plant configuration. Systems engineering is thus optimized, reducing dependence on electrical and instrumentation and control (I and C) systems. The fuel assemblies deployed in the SWR 1000 core are enlarged from a 10 x 10 to a 12 x 12 rod array. This cuts down the total number of fuel assemblies in the core and hence also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment, maintenance and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope for outage activities. For example, there are no bolted reactor internal joints, and the SWR 1000 has a permanently installed reactor vessel-to-drywell seal. Replacement of in-core detectors is carried out from below, and does not affect the critical path of the outage. Furthermore all in-service -inspections (ISIs) on the reactor pressure vessel (RPV) and its nozzles can be undertaken from the outside, so that no extended ISI outages are necessary. As in existing plants, forced coolant circulation is utilized, ensuring problem-free startup and enabling plant operators to adjust power rapidly in the high power range (70%-100%). This is achieved without moving the control rods, and allows both spectral-shift and stretch-out operation. The considerable gains provided by forced coolant circulation, in terms of operational flexibility and fuel utilization, mean that the investment and maintenance costs of the pumps are covered within just a few

DVL Velocity Aiding in the HUGIN 1000 Integrated Inertial Navigation System

Directory of Open Access Journals (Sweden)

Bjørn Jalving

2004-10-01

Full Text Available The RDI WHN-600 Doppler Velocity Log (DVL is a key navigation sensor for the HUG1N 1000 Autonomous Underwater Vehicle (AUV. HUGIN 1000 is designed for autonomous submerged operation for long periods of time. This is facilitated by a low drift velocity aided Inertial Navigation System (INS. Major factors determining the position error growth are the IMU and DVL error characteristics and the mission plan pattern_ For instance, low frequency DVL errors cause an approximately linear drift in a straight-line trajectory, while these errors tend to be cancelled out by a lawn mower pattern_ The paper focuses on the accuracy offered by the DVL. HUGIN 1000 is a permanent organic mine countermeasure (MCM capacity on the Royal Norwegian Navy MCM vessel KNM Karmoy. HUGIN 1000 will be part of the NATO force MCMFORNORTH in fall 2004.

The modularization construction of piping system installation in AP1000 plant

International Nuclear Information System (INIS)

Lu Song; Wang Yuan; Wei Junming

2012-01-01

Modularization construction is the main technique used in AP1000 plants, the piping Modularization installation will impact directly to the module construction as the important part of the Modularization construction. After the piping system has took the modularization design in AP1000 plants, some installation works of piping system has moved from the site to fabrication shop. With improving the construction quality and minimizing the time frame of project, the critical paths can be optimized. This paper has analyzed the risk and challenge that met during the modularization construction period of piping systems though introducing the characteristic of modularization construction for AP1000 piping systems, and get construction experiences from the First AP1000 plants in the world, then it will be the firmly basics for the wide application of modularization construction in the future. (authors)

A study on the economics enhancement of OPR1000 applied to advanced construction methods

International Nuclear Information System (INIS)

Park, Ki Jo; Yoon, Eun Sang

2007-01-01

OPR1000 (Optimized Power Reactor 1000MW) is a totally improved design model of Korea nuclear power plants and the latest 1,000MW nuclear power plant in the Republic of Korea. Shin Kori 1 and 2 and Shin Wolsong 1 and 2 and under construction and these are OPR1000 types. Although OPR1000 is up to data 1,000MW nuclear power plant, it is not enough to be much superior to other nuclear power plants. Under the WTO and FTA circumstance of domestic and stiff overseas competition for nuclear power plants, it is necessary to enhance the economics of OPR1000. And then, the enhanced economic alternatives are reviewed and the advanced construction methods are considered. Based on research and a comprehensive review of nuclear power plant construction experiences, an alternative application of advanced construction methods is developed and compared with existing OPR1000 for schedule and economics. In this paper, economic analyses of a construction cost and a levelized electricity generation cost are performed

Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

Energy Technology Data Exchange (ETDEWEB)

Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

2012-10-15

The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

24 CFR 1000.14 - What relocation and real property acquisition policies are applicable?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What relocation and real property acquisition policies are applicable? 1000.14 Section 1000.14 Housing and Urban Development Regulations... HOUSING, DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.14...

24 CFR 1000.50 - What Indian preference requirements apply to IHBG administration activities?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What Indian preference requirements apply to IHBG administration activities? 1000.50 Section 1000.50 Housing and Urban Development... § 1000.50 What Indian preference requirements apply to IHBG administration activities? To the greatest...

Dynamic Analysis of AP1000 Shield Building Considering Fluid and Structure Interaction Effects

Directory of Open Access Journals (Sweden)

Qiang Xu

2016-02-01

Full Text Available The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH and finite element method (FEM coupling method is used to numerically simulate the fluid and structure interaction (FSI between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

Hbo's: 1000 euro extra per allochtone leerling

NARCIS (Netherlands)

De Volkskrant

2006-01-01

De vier grote hogescholen in de Randstad willen 1000 euro extra voor elke student van laag opgeleide ouders. Bestuursvoorzitter Pim Breebaart van de Haag-se Hogeschool verwacht dat dit de schatkist bijna 20 miljoen euro zal kosten.

24 CFR 1000.56 - How are NAHASDA funds paid by HUD to recipients?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false How are NAHASDA funds paid by HUD to recipients? 1000.56 Section 1000.56 Housing and Urban Development Regulations Relating to Housing..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.56 How are...

24 CFR 1000.1 - What is the applicability and scope of these regulations?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What is the applicability and scope of these regulations? 1000.1 Section 1000.1 Housing and Urban Development Regulations Relating to..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.1 What is the...

Comparison and analysis for item classifications between AP1000 and traditional PWR

International Nuclear Information System (INIS)

Luo Shuiyun; Liu Xiaoyan

2012-01-01

The comparison and analysis for the safety classification, seismic category, code classification and QA classification between AP1000 and traditional PWR were presented. The safety could be guaranteed and the construction and manufacture costs could be cut down since all sorts of AP1000 classifications. It is suggested that the QA classification and the QA requirements correspond to the national conditions should be drafted in the process of AP1000 domestication. (authors)

Introduction to sump screen downstream effect analysis of AP1000 nuclear power plant

International Nuclear Information System (INIS)

Zhang Qinghua; Liu Yu; Chai Guohan

2010-01-01

The design of AP1000 takes into account the potential impact of debris clogging on sump screen. In this article, the technical background of sump screen issue and the design characteristics of AP1000 to address the sump screen blockage issue are introduced. The article focuses on the 'downstream effect' analysis method, acceptance criteria and analysis result of AP1000 sump screen. Although the design of AP1000 is different with traditional PWR, the author expects to bring some reference to advance the downstream effect analysis in China through the introduction. (authors)

Cleaning of OPR1000 Steam Generator by Ultrasonic Cavitation in Water

Energy Technology Data Exchange (ETDEWEB)

Jeong, Wootae [Korea Hydro and Nuclear Power Co., Ltd, Daejeon (Korea, Republic of); Kim, Sangtae; Yoon, Sangjung; Choi, Yongseok [Saean Engineering Corporation, Seoul (Korea, Republic of)

2013-05-15

Magnetic wheels are attached to the transducers to prevent tube damage which may be caused by wear between the transducers and SG tubes. To remove heat generated by transducers, we used water to water heat exchanger. Sludge removed from tube sheet area of the steam generator was pumped to filtering station for removing impurities in it. We designed an ultrasonic cleaning system for application to OPR1000 S/G. The technology was developed for removing sludge in OPR1000 S/G. However, the technology could easily be applied to other types of S/Gs. For cleaning OPR1000 SG, we designed an ultrasonic cleaning system with 12 transducers, 15 generators, a WRS, and a water treatment system. An experiment with a single transducer and the full scale OPR1000 S/G mock-up did not show very satisfactory result in ultrasound energy level. However, we expect sufficient effects if we apply 12 or more transducers in this case considering our previous experimental results as shown in the references. The ultrasonic cleaning system will be ready in August this year for performance test. After several experiments and the experiments followed, we are planning to apply this cleaning system for removing sludge in Korean OPR1000 S/Gs.

24 CFR 1000.126 - May a recipient charge flat or income-adjusted rents?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false May a recipient charge flat or income-adjusted rents? 1000.126 Section 1000.126 Housing and Urban Development Regulations Relating to... § 1000.126 May a recipient charge flat or income-adjusted rents? Yes, providing the rental or homebuyer...

VVER-1000: considering its strengths and weaknesses

International Nuclear Information System (INIS)

Laaksonen, J.

1994-01-01

The safety of currently operating VVER-1000 reactors is examined. The factors considered are deviations in operation, inherent safety, safety system design, protection against internal and external hazards, equipment quality, the approach to plant operations and the safety culture. On the basis of this evaluation it is concluded that the overall safety of a VVER-1000 cannot be at the level of a modern Western PWR though there is no sound basis to make a quantitative comparison. Many of the concerns raised are being adequately addressed in the Czech Temelin which is currently under construction and in new designs which are still at the drawing board stage. Extensive back fitting programmes are planned or underway in operating plants. The creation of independent responsible operating organizations, powerful regulation and an improved economic situation are advanced as necessary criteria for real improvements in safety. (UK)

Reliability of the Nidek NT-1000 non contact tonometer.

Science.gov (United States)

Van de Velde, T; Zeyen, T

1995-01-01

In this prospective study, we compared the intraocular pressure (IOP) readings of 100 patients measured with the Goldmann applanation tonometer and the Nidek NT-1000 pneumotonometer. The correlation coefficient between the Goldmann and Nidek readings was 0.86. On the average the pneumotonometer overestimated the intraocular pressure with 0.43 mm Hg. The Nidek NT-1000 non contact tonometer can be used for screening purposes provided an appropriately low IOP value is used to indicate the need for further assessment with the Goldman applanation tonometer.

AP1000{sup R} severe accident features and post-Fukushima considerations

Energy Technology Data Exchange (ETDEWEB)

Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

Evaluation of the MIT RMID 1000 system for the identification of Listeria species.

Science.gov (United States)

Ricardi, John; Haavig, David; Cruz, Lasaunta; Paoli, George; Gehring, Andrew

2010-01-01

The Micro Imaging Technology (MIT) 1000 Rapid Microbial Identification (RMID) System is a device that uses the principles of light scattering coupled with proprietary algorithms to identify bacteria after being cultured and placed in a vial of filtered water. This specific method is for pure culture identification of Listeria spp. A total of 81 microorganisms (55 isolates) were tested by the MIT 1000 System, of which 25 were Listeria spp. and 30 a variety of other bacterial species. In addition, a total of 406 tests over seven different ruggedness parameters were tested by the MIT 1000 System to determine its flexibility to the specifications stated in the MIT 1000 System User Guide in areas where they might be deviated by a user to shorten the test cycle. Overall, MIT concluded that the MIT 1000 System had an accuracy performance that should certify this Performance Test Method for the identification of Listeria spp. This report discusses the tests performed, results achieved, and conclusions, along with several reference documents to enable a higher understanding of the technology used by the MIT 1000 System.

The Westinghouse Series 1000 Mobile Phone: Technology and applications

Science.gov (United States)

Connelly, Brian

1993-01-01

Mobile satellite communications will be popularized by the North American Mobile Satellite (MSAT) system. The success of the overall system is dependent upon the quality of the mobile units. Westinghouse is designing our unit, the Series 1000 Mobile Phone, with the user in mind. The architecture and technology aim at providing optimum performance at a low per unit cost. The features and functions of the Series 1000 Mobile Phone have been defined by potential MSAT users. The latter portion of this paper deals with who those users may be.

24 CFR 1000.106 - What families receiving assistance under title II of NAHASDA require HUD approval?

Science.gov (United States)

2010-04-01

... under title II of NAHASDA require HUD approval? 1000.106 Section 1000.106 Housing and Urban Development... Housing Activities § 1000.106 What families receiving assistance under title II of NAHASDA require HUD... in §§ 1000.108 and 1000.110. (b) Assistance under section 201(b)(3) of NAHASDA for non-Indian...

Economic impact evaluation of the Procap 1000: Deep Water Qualification Program from PETROBRAS

International Nuclear Information System (INIS)

Furtado, Andre T.; Pereira, Newton M.; Suslick, Saul; Freitas, Adriana G. de; Bach, Laurent

1999-01-01

PETROBRAS, a Brazilian petroleum company, managed between 1986 and 1992 a program with purpose to dominate the necessary technology for the petroleum production up to 1000 meters depth. This program was going called Procap 1000. The aim of the work was to evaluate the impacts of Procap 1000. The proposed evaluation method by Beta was going used. The results are presented

Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

Directory of Open Access Journals (Sweden)

Voloschenko Andrey

2016-01-01

Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

Standardized safety management of AP1000 nuclear power plant

International Nuclear Information System (INIS)

Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

2011-01-01

In 2002, China published and implemented the Law of the People's Republic of China on Work Safety and promulgated a series of guidelines and policies, which strengthened the safety management supervision. Standardization of safety, as another important step on safety supervision, comes after safety assesment and safety production licensing system, is also a permanent solution. Standardization of safety is a strategic, long term and fundamental work, which is also the basic access to achieving scientific safety management and increasing the inherent safety of an enterprise. Haiyang AP1000 nuclear power plant, adopting the modularized, 'open-top' and parallel construction means, overturned the traditional construction theory of installation work comes after the civil work and greatly shorten the construction period. At the same time, the notable increase of oversize module transportation and lifting and parallel construction raises higher demands for safety management. This article combines the characteristics and difficulties of safety management for Haiyang AP1000 nuclear power plant, puts forward ideas and methods for standardized safety management, and could also serve as reference to the safety management for other AP1000 projects. (authors)

Steam turbine of WWER-1000 unit

International Nuclear Information System (INIS)

Drahy, J.

1986-01-01

The manufacture was started by Skoda of a saturated steam, 1,000 MW, 3,000 rpm turbine designed for the Temelin nuclear power plant. The turbine provides steam for heating water for district heating, this either with an output of 893 MW for a three-stage water heating at 150/60 degC, or of 570 MW for a two-stage water heating at 120/60 degC. The turbine features one high-pressure and three identical low-pressure stages. The pressure gradient between the high-pressure and the low-pressure parts was optimized with respect to the thermal efficiency of the cycle and to the thermodynamic efficiency of the low-pressure part. A value of 0.79 MPa was selected corresponding to the maximum through-flow of steam entering the turbine. This makes 5,495 t/h, the admission steam parameters are 273.3 degC and 5.8 MPa. The feed water temperature is 220.9 degC. 300 cold starts, 1,000 starts after shutdowns for 55 to 88 hours and 600 starts after shutdown for 8 hours are envisaged for the entire turbine service life. (Z.M.). 5 figs., 1 tab., 6 refs

Control parameter optimization for AP1000 reactor using Particle Swarm Optimization

International Nuclear Information System (INIS)

Wang, Pengfei; Wan, Jiashuang; Luo, Run; Zhao, Fuyu; Wei, Xinyu

2016-01-01

Highlights: • The PSO algorithm is applied for control parameter optimization of AP1000 reactor. • Key parameters of the MSHIM control system are optimized. • Optimization results are evaluated though simulations and quantitative analysis. - Abstract: The advanced mechanical shim (MSHIM) core control strategy is implemented in the AP1000 reactor for core reactivity and axial power distribution control simultaneously. The MSHIM core control system can provide superior reactor control capabilities via automatic rod control only. This enables the AP1000 to perform power change operations automatically without the soluble boron concentration adjustments. In this paper, the Particle Swarm Optimization (PSO) algorithm has been applied for the parameter optimization of the MSHIM control system to acquire better reactor control performance for AP1000. System requirements such as power control performance, control bank movement and AO control constraints are reflected in the objective function. Dynamic simulations are performed based on an AP1000 reactor simulation platform in each iteration of the optimization process to calculate the fitness values of particles in the swarm. The simulation platform is developed in Matlab/Simulink environment with implementation of a nodal core model and the MSHIM control strategy. Based on the simulation platform, the typical 10% step load decrease transient from 100% to 90% full power is simulated and the objective function used for control parameter tuning is directly incorporated in the simulation results. With successful implementation of the PSO algorithm in the control parameter optimization of AP1000 reactor, four key parameters of the MSHIM control system are optimized. It has been demonstrated by the calculation results that the optimized MSHIM control system parameters can improve the reactor power control capability and reduce the control rod movement without compromising AO control. Therefore, the PSO based optimization

Investigation of the Effective Thermal Conductivity in Containment Wall of OPR1000

Energy Technology Data Exchange (ETDEWEB)

Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Lee, Jong Hwi; Kang, Hie Chan [Kunsan National University, Gunsan (Korea, Republic of)

2016-05-15

Many computational codes used for analyzing pressure of containment was developed such as CAP (Containment Analysis Package). These computational codes consider concrete conductivity instead of thermal conductivity of containment wall which have special geometry as heat sink. For precise analysis, effective thermal conductivity of containment wall has to be measured in individual NPPs. Thermal properties of concrete such as thermal conductivity have been investigated as function of chemical composition and temperature. Generally, containment of OPR1000 is constructed by Prestressed (PS) concrete-a composite material. Containment wall of OPR1000 is made up of steel liner, tendon, rebar and concrete as shown in Figure 1. Role of steel liner protects release of radioactive materials so called leak tightness. The effective thermal conductivity of containment wall in OPR1000 is analyzed by numerical tool (CFD) and compared with thermal conductivity models in composite solids. The effective thermal conductivity of containment wall of OPR1000 is investigated by numerical analysis (CFD). The thermal conductivity of reinforced concrete is 18.6% higher than that of concrete only. Several models were compared with CFD results. Rayleigh-Parallel liner model agrees well with CFD results. Experiment results will be compared with CFD result and models. CFD result was calculated in low steel volume fraction (0.0809) than that of OPR1000 (0.1043). The effective thermal conductivity in OPR1000 has slightly higher than CFD result because of different volume fraction.

1000th magnet delivered!

CERN Multimedia

2006-01-01

On Monday 20 February members of the AT Department marked the delivery of the 1000th superconducting dipole magnet to CERN. Only 232 more of the dipole magnets are needed for the LHC. The 35 tonne-dipoles are 15 meters long and are being manufactured by three companies: Babcock Noell Nuclear in Germany (which finished its contract in November 2005), Ansaldo Superconduttori in Italy and Alstom-Jeumont in France. "The production is proceeding well and we expect to be complete in October as previously foreseen," said Lucio Rossi, Head of the Magnets and Superconductors Group (AT-MAS). In total, 1650 main magnets are needed for the LHC, of which 1300 have been delivered.

1000th magnet delivered!

CERN Multimedia

2006-01-01

On Monday 20 February members of the AT Department marked the delivery of the 1000th superconducting dipole magnet to CERN. Only 232 more of the dipole magnets are needed for the LHC. The 35-tonne-dipoles are 15 meters long and are being manufactured by three companies: Babcock Noell Nuclear in Germany (which completed its contract in November 2005), Ansaldo Superconduttori in Italy and Alstom-Jeumont in France. 'The production is proceeding well and we expect to be complete in October as foreseen,' said Lucio Rossi, Head of the Magnets and Superconductors Group (AT-MAS). In total, 1650 main magnets are needed for the LHC, of which 1300 have already been delivered.

Structural modules in AP1000 plant design

International Nuclear Information System (INIS)

Prasad, N.; Tunon-Sanjur, L.

2007-01-01

Structural modules are extensively used in AP1000 plant design. The shop manufacturing of modules components improves the quality and reliability of plant structures. The application of modules has a positive impact on construction schedules, and results in substantial savings in the construction cost. This paper describes various types of structural modules used for AP1000 plant structures. CA structural wall modules are steel plate modules with concrete placed, on or within the module, after module installation. The layout and design of the largest CA wall modules, CA01 and CA20, is described in detail. General discussion of structural floor modules, such as the composite and finned floors, is also included. Steel form CB modules (liners) consist of plate reinforced with angle stiffeners and tee sections. The angles and the tee sections are on the concrete side of the plate. Design of CB20 has been included as an example of CB type modules. Design codes and structural concepts related to module designs are discussed. (authors)

VVER-1000 RPV Head Examination Control System

International Nuclear Information System (INIS)

Erak, Z.; Gortan, K.

2006-01-01

This article presents the electronic system used for automated NDT examination of VVER-1000 Reactor Pressure Vessel Head (RPVH). The control system drives the inspection tool with end-effectors to needed position. When the final position is reached, the eddy current and ultra sound acquisition system performs the data acquisition. The system is composed of 3 layers. The first layer is the hardware layer consisting of motors driving the tool and end-effectors along with sensors needed to obtain the positioning data. The second layer is the MAC-8 control system performing basic monitoring and control routines as an interconnection between first and third layer. The third layer is the control software, running on PC, which is used as a human-machine-interface. Presentation contains details of examination techniques with focus on eddy current examination as well as details on manipulator and end effectors developed by Inetec for VVER-1000 RPVH examination.(author)

24 CFR 1000.2 - What are the guiding principles in the implementation of NAHASDA?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What are the guiding principles in the implementation of NAHASDA? 1000.2 Section 1000.2 Housing and Urban Development Regulations... HOUSING, DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.2...

Development of TVSA VVER-1000 fuel

International Nuclear Information System (INIS)

Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

2013-01-01

The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

Vähemalt 1000-euroste arvete deklareerimine / Tõnis Elling

Index Scriptorium Estoniae

Elling, Tõnis, 1972-

2014-01-01

1. novembril 2014 jõustuvast käibemaksuseaduse muudatusest, mille kohaselt peavad kõik käibemaksukohustuslased hakkama käibedeklaratsiooni (KMD) lisal INF deklareerima vähemalt 1000 euro suuruseid arveid

24 CFR 1000.20 - Is an Indian tribe required to assume environmental review responsibilities?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Is an Indian tribe required to assume environmental review responsibilities? 1000.20 Section 1000.20 Housing and Urban Development... § 1000.20 Is an Indian tribe required to assume environmental review responsibilities? (a) No. It is an...

24 CFR 1000.112 - How will HUD determine whether to approve model housing activities?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false How will HUD determine whether to approve model housing activities? 1000.112 Section 1000.112 Housing and Urban Development Regulations... Activities § 1000.112 How will HUD determine whether to approve model housing activities? HUD will review all...

24 CFR 1000.22 - Are the costs of the environmental review an eligible cost?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Are the costs of the environmental review an eligible cost? 1000.22 Section 1000.22 Housing and Urban Development Regulations Relating to..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.22 Are the...

Project W-320, backup: 1000 CFM portable exhausters acceptance for beneficial use

International Nuclear Information System (INIS)

Nelson, O.D.

1998-01-01

This document is to identify the Project W-320 1000 CFM portable exhauster documentation required to be turned over from the Projects Organization to the Tank Farm Operations as part of the acceptance of the 1000 CFM portable exhausters for beneficial use

Nuclear power plants with reactors WWER-1000 type: today and tomorrow; AEhS s WWER-1000: nastoyashchee i budushchee

Energy Technology Data Exchange (ETDEWEB)

Molchanov, V; Biryukov, G; Novak, K [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

1996-12-31

There are currently 19 NPP units based on WWER-1000 reactors working in Russia, Ukraine and Bulgaria. They are of four types: V-187, V-302, V-338, V-320. The design principles of these reactors comply with regulations of the eighties, and it is necessary to introduce improvements according to the new regulations and to the operation experience gained. Two approaches for safety and efficiency enhancement are described: AS-91 and AS-92. AS-91 implies gradual improvement of the base WWER-1000/V-320 design by incorporation of new design solutions avoiding the need of building large scale models. AS-92 refers to entirely new design which require experimental research by building a full scale models or by using natural stands. The latter approach will be used for NPP projects to be built after year 2000. The main new feature of AS-92 is the addition of passive safety systems to the active ones in order to protect the fuel from damage.

Estimation of material degradation of VVER-1000 baffle

Science.gov (United States)

Harutyunyan, Davit; Koš'ál, Michal; Vandlík, Stanislav; Hojná, Anna; Schulc, Martin; Flibor, Stanislav

2017-09-01

The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

Characteristics and application study of AP1000 NPPs equipment reliability classification method

International Nuclear Information System (INIS)

Guan Gao

2013-01-01

AP1000 nuclear power plant applies an integrated approach to establish equipment reliability classification, which includes probabilistic risk assessment technique, maintenance rule administrative, power production reliability classification and functional equipment group bounding method, and eventually classify equipment reliability into 4 levels. This classification process and result are very different from classical RCM and streamlined RCM. It studied the characteristic of AP1000 equipment reliability classification approach, considered that equipment reliability classification should effectively support maintenance strategy development and work process control, recommended to use a combined RCM method to establish the future equipment reliability program of AP1000 nuclear power plants. (authors)

Preparing for 1000 GeV physics at Fermilab

International Nuclear Information System (INIS)

Anon.

1980-01-01

The superconducting proton beams and the neutrino beams at Fermilab prepared for the research with 1000 GeV colliding proton and antiproton beams are described. Especially a new developed helium transfer line is described. (HSI).

OLP embedment design method research for AP1000 nuclear plant

International Nuclear Information System (INIS)

Li Cheng; Li Shaoping; Liu Jianwei

2013-01-01

Background: One of the most advanced nuclear power technology, the first AP1000 reactor is under construction in China. Modularization is one of the main characteristics for AP1000 nuclear plant building. Module wall with steel face plate is used instead of reinforced concrete structure wall. A number of OLP embedments need to be installed into the module wall to connect other structures such as pipes, equipment, operation platforms and any other component attached to the module wall. Therefore, the design of embedment is very important in AP1000 structural design. Purpose: A finite element analysis method and tool for embedment design is needed for convenience. Methods: This paper applies the self-developed GTStrudl command template and VBA macro program for embedment capacity calculation and evaluation based on Microsoft Excel to the embedment design. Results: A Microsoft Excel template for embedment design is developed. Conclusions: The analysis method and template brings reasonable results and may provide some help and use for reference for the engineering practice. (authors)

24 CFR 1000.60 - Can HUD prevent improper expenditure of funds already disbursed to a recipient?

Science.gov (United States)

2010-04-01

... expenditure of funds already disbursed to a recipient? 1000.60 Section 1000.60 Housing and Urban Development... § 1000.60 Can HUD prevent improper expenditure of funds already disbursed to a recipient? Yes. In accordance with the standards and remedies contained in § 1000.538 relating to substantial noncompliance, HUD...

Remote technology in RBMK-1000 spent fuel management at NPP site

International Nuclear Information System (INIS)

Makarchuk, T.F.; Kozlov, Y.V.; Tikhonov, N.S.; Tokarenko, A.I.; Spichev, V.V.; Kaljazin, N.N.

1999-01-01

The report describes the remote technologies employed in the nuclear power plant with RBMK-1000 type. Spent fuel transfer and handling operations at reactor (AR) and away from reactor (AFR) on reactor site (RS) facilities are illustrated by the example of the Leningradskaya NPP and are typical for all NPPs with RBMK-1000. The current approach to spent fuel management at NPP sites is also presented. (author)

7 CFR 1000.43 - General classification rules.

Science.gov (United States)

2010-01-01

... market administrator shall determine the shrinkage or overage of skim milk and butterfat for each pool... pursuant to § 1000.44, the following rules shall apply: (a) Each month the market administrator shall...) and § 1135.11 of this chapter, the pounds of skim milk and butterfat, respectively, in each class in...

Modelling of WWER-1000 fuel: state and prospects

International Nuclear Information System (INIS)

Medvedev, A.; Bibilashvili, Yu.; Bogatyr, S.; Khvostov, G.

1994-01-01

The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs

Modelling of WWER-1000 fuel: state and prospects

Energy Technology Data Exchange (ETDEWEB)

Medvedev, A; Bibilashvili, Yu; Bogatyr, S; Khvostov, G [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

1994-12-31

The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs.

The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

International Nuclear Information System (INIS)

Demetri, K. J.; Leipner, C. I.; Marshall, M. L.

2015-09-01

The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

Energy Technology Data Exchange (ETDEWEB)

Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2015-09-15

The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

PEGylated DX-1000: Pharmacokinetics and Antineoplastic Activity of a Specific Plasmin Inhibitor

Directory of Open Access Journals (Sweden)

Laetitia Devy

2007-11-01

Full Text Available Novel inhibitors of the urokinase-mediated plasminogen (plg activation system are potentially of great clinical benefit as anticancer treatments. Using phage display, we identified DX-1000 a tissue factor pathway inhibitor-derived Kunitz domain protein which is a specific high-affinity inhibitor of plasmin (pin (Ki = 99 pM. When tested in vitro, DX-1000 blocks plasminmediated pro-matrix metal loproteinase-9 (proMMP-9 activation on cells and dose-dependently inhibits tube formation, while not significantly affecting hemostasis and coagulation. However, this low-molecular weight protein inhibitor (~ 7 kDa exhibits rapid plasma clearance in mice and rabbits, limiting its potential clinical use in chronic diseases. After site-specific PEGylation, DX-1000 retains its activity and exhibits a decreased plasma clearance. This PEGylated derivative is effective in vitro, as well as potent in inhibiting tumor growth of green fluorescent protein (GFP-labeled MDA-MB-231 cells. 4PEG-DX-1000 treatment causes a significant reduction of urokinase-type plasminogen activator (uPA and plasminogen expressions, a reduction of tumor proliferation, and vascularization. 4PEG-DX-1000 treatment significantly decreases the level of active mitogenactivated protein kinase (MAPK in the primary tumors and reduces metastasis incidence. Together, our results demonstrate the potential value of plasmin inhibitors as therapeutic agents for blocking breast cancer growth and metastasis.

Study on the operation mode for indigenization and standardization of AP1000 technology

International Nuclear Information System (INIS)

Gao Zhihu; Cheng Huiping

2014-01-01

This paper describes the importance and necessity of developing standardized AP1000 technology, and analyzes the problems faced and measures to be taken. The operation mode, known as the Committee of AP1000 Standardized Design and Innovation Management, was first put forward in China. And the paper also discusses how to arouse the enthusiasm of the owners, designers, manufactures, as well as construction, installation, regulation and other parties, how to exploit the advantage of whole industry to promote the localization and standardization of AP1000 technology. (authors)

The study on neutron and photon distribution of AP1000 reactor by MCNP code

International Nuclear Information System (INIS)

Chen Defeng; Shen Mingqi

2014-01-01

The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

25 CFR 1000.223 - When can a Tribe/Consortium request a waiver of a regulation?

Science.gov (United States)

2010-04-01

... regulation? 1000.223 Section 1000.223 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Waiver of Regulations § 1000.223 When can a Tribe/Consortium...

25 CFR 1000.162 - What is included in a self-governance compact?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What is included in a self-governance compact? 1000.162...-DETERMINATION AND EDUCATION ACT Negotiation Process for Annual Funding Agreements Negotiating A Self-Governance Compact § 1000.162 What is included in a self-governance compact? A model format for self-governance...

Comparison of HapMap and 1000 Genomes Reference Panels in a Large-Scale Genome-Wide Association Study.

Directory of Open Access Journals (Sweden)

Paul S de Vries

Full Text Available An increasing number of genome-wide association (GWA studies are now using the higher resolution 1000 Genomes Project reference panel (1000G for imputation, with the expectation that 1000G imputation will lead to the discovery of additional associated loci when compared to HapMap imputation. In order to assess the improvement of 1000G over HapMap imputation in identifying associated loci, we compared the results of GWA studies of circulating fibrinogen based on the two reference panels. Using both HapMap and 1000G imputation we performed a meta-analysis of 22 studies comprising the same 91,953 individuals. We identified six additional signals using 1000G imputation, while 29 loci were associated using both HapMap and 1000G imputation. One locus identified using HapMap imputation was not significant using 1000G imputation. The genome-wide significance threshold of 5×10-8 is based on the number of independent statistical tests using HapMap imputation, and 1000G imputation may lead to further independent tests that should be corrected for. When using a stricter Bonferroni correction for the 1000G GWA study (P-value < 2.5×10-8, the number of loci significant only using HapMap imputation increased to 4 while the number of loci significant only using 1000G decreased to 5. In conclusion, 1000G imputation enabled the identification of 20% more loci than HapMap imputation, although the advantage of 1000G imputation became less clear when a stricter Bonferroni correction was used. More generally, our results provide insights that are applicable to the implementation of other dense reference panels that are under development.

Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

International Nuclear Information System (INIS)

Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

2011-01-01

Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

Operational safety improvement in OPR 1000

International Nuclear Information System (INIS)

Jung, Y.-E.

2005-01-01

Nuclear power operating experience management might be an important factor for the operational safety improvement. KHNP's nuclear information management system, called KONIS receives, distributes and manages all nuclear information from domestic and foreign, especially operating experience. Ulchin 3 and 4, the first units of OPR 1000 series operates several organizations regarding management of operating experience e.g. specialist group program, various task forces, equipment specialist system for operator, etc. Peer review is another contribution for nuclear safety. (author)

25 CFR 1000.166 - What are the phases of the negotiation process?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What are the phases of the negotiation process? 1000.166 Section 1000.166 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Negotiation Process...

ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-06-01

Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

Validation of MCNP and WIMS-AECL/DRAGON/RFSP for ACR-1000 applications

International Nuclear Information System (INIS)

Bromley, Blair P.; Adams, Fred P.; Zeller, Michael B.; Watts, David G.; Shukhman, Boris V.; Pencer, Jeremy

2008-01-01

This paper gives a summary of the validation of the reactor physics codes WIMS-AECL, DRAGON, RFSP and MCNP5, which are being used in the design, operation, and safety analysis of the ACR-1000 R . The standards and guidelines being followed for code validation of the suite are established in CSA Standard N286.7-99 and ANS Standard ANS-19.3-2005. These codes are being validated for the calculation of key output parameters associated with various reactor physics phenomena of importance during normal operations and postulated accident conditions in an ACR-1000 reactor. Experimental data from a variety of sources are being used for validation. The bulk of the validation data is from critical experiments in the ZED-2 research reactor with ACR-type lattices. To supplement and complement ZED-2 data, qualified and applicable data are being taken from other power and research reactors, such as existing CANDU R units, FUGEN, NRU and SPERT research reactors, and the DCA critical facility. MCNP simulations of the ACR-1000 are also being used for validating WIMS-AECL/ DRAGON/RFSP, which involves extending the validation results for MCNP through the assistance of TSUNAMI analyses. Code validation against commissioning data in the first-build ACR-1000 will be confirmatory. The code validation is establishing the biases and uncertainties in the calculations of the WIMS-AECL/DRAGON/RFSP suite for the evaluation of various key parameters of importance in the reactor physics analysis of the ACR-1000. (authors)

Extended analysis of WWER-1000 Charpy test data

International Nuclear Information System (INIS)

Vodenicharov, St.; Kamenova, Tz.

2001-01-01

The aim of this work is to study the embrittlement rate of WWER-1000 RPV weld metal with high Ni content and to determine influence of neutron irradiation on partial energies of ductile crack initiation, stable and unstable crack propagation and post crack arrest. (author)

Oregon state university's advanced plant experiment (APEX) AP1000 integral facility test program

International Nuclear Information System (INIS)

Reyes, J.N.; Groome, J.T.; Woods, B.G.; Young, E.; Abel, K.; Wu, Q.

2005-01-01

Oregon State University (OSU) has recently completed a three year study of the thermal hydraulic behavior of the Westinghouse AP1000 passive safety systems. Eleven Design Basis Accident (DBA) scenarios, sponsored by the U.S. Department of Energy (DOE) with technical support from Westinghouse Electric, were simulated in OSU's Advanced Plant Experiment (APEX)-1000. The OSU test program was conducted within the purview of the requirements of 10CFR50 Appendix B, NQA-1 and 10 CFR 21 and the test data was used to provide benchmarks for computer codes used in the final design approval of the AP1000. In addition to the DOE certification testing, OSU conducted eleven confirmatory tests for the U.S. Nuclear Regulatory Commission. This paper presents the test program objectives, a description of the APEX-1000 test facility and an overview of the test matrix that was conducted in support of plant certification. (authors)

An investigation of axial xenon stability in WWER-1000 reactor designs

International Nuclear Information System (INIS)

Doshi, P.K.; Miller, R.W.

1993-01-01

The nuclear power plants of the WWER-1000 design have experienced frequent xenon oscillation control problems. In most PWRs, xenon oscillations are largely a problem in the axial direction. An one dimensional core model representative of the WWER-1000 design was set up to examine the controllability of the current design. An investigation of possible improvements to this design was made. There was no indication that xenon oscillations were an inherent problem in WWER-1000 core design. Simple changes to the control rod system coupled with a sound power distribution control strategy that has been proven to be an effective but simple procedure to follow, eliminate xenon control problems. The changes proposed can be implemented in a very cost effective manner. There are no equipment changes needed, existing control rods can be used. Only software changes are required. (Z.S.) 1 tab., 2 figs., 7 refs

The economic and financing viability of the OPR-1000 construction at Muria Peninsula

International Nuclear Information System (INIS)

Moch Djoko Birmano; Mochamad Nasrullah

2005-01-01

The study of the economic and financing viability of the OPR-1000 construction at Muria Peninsula have been done. This study is carried out with steps as follows: updating of newest OPR-1000 technical and economic data; survey of financing sources; the calculation of electricity generation cost, electricity tariff, construction cost of OPR-1000; and calculation of OPR-1000 financing viability by using feasibility criteria such as Net Present Value (NPV), Internal Rate of Return (IRR) and capital payback period. The calculation of generation cost, electricity tariff, construction cost and financing viability criteria (IRR, NPV, Payback Period)by using KEPCO Spread Sheet. From the results of calculations how that the electricity generation cost is 4.0866 cent/kWh, while the electricity tariff is 6.6399 cent/kWh(after Value Added Tax). The total construction cost of OPR-1000 is US $4,092.09 million (included interest and financial fee, excluded initial nuclear fuel). The calculation of financing viability in base case obtained that value of IRR, NPV and Payback Period for total investment is 10.37%. US$90.52 million and 12.11 years respectively. From these results can be concluded that with the electricity tariff of 6.640 cent/kWh or 0.0664 US$/kWh, basically this project of OPR-1000 construction is very feasible and beneficial. From investment side, this project can attract investor because the rate of capital return is high enough, profit in the end of economic life is big enough and payback period is short. (author)

Usefulness of 1000-Hz probe tone in tympanometry according to age in Korean infants.

Science.gov (United States)

Park, Mina; Han, Kyu-Hee; Jung, Hyunseo; Kim, Mee-Hee; Chang, Hyun-Kyung; Kim, Shin Hye; Park, Moo Kyun; Lee, Jun Ho

2015-01-01

Numerous studies have shown the superiority of a 1000-Hz frequency probe tone for evaluating the middle ear status of infants. However, most of these studies examined Caucasian populations. This study validated the 1000-Hz probe tone and evaluated the age at which it should be used in Korean infants. Data from 83 infants (43 males, 40 females; mean age 9.2±6.2 (range 1-30) months, 165 ears) were analyzed. Tympanograms were classified according to Baldwin's modification of the method of Marchant et al. and correlated with results based on combined diagnostic tests, including an endoscopic examination of the tympanic membrane, myringotomy findings, and the air and bone conduction auditory brainstem response (ABR) thresholds. Data were analyzed in five age groups, each covering a 3-month range. The traces were measured for both 226- and 1000-Hz probe tones. The sensitivity and specificity for the different age groups were also determined. For the 226-Hz probe tone, the tympanograms showed normal traces for most ears with otitis media effusions in infants younger than 12 months. By contrast, the tympanograms using the 1000-Hz probe tone showed abnormal traces in most of the infants with otitis media effusions in all age groups. In infants with no otitis media effusion, the tympanograms using both 226- and 1000-Hz probe tones were interpreted as normal in most cases in all age groups. In infants younger than 12 months, the sensitivity of the 226-Hz probe tone was very low (0-6.6%), whereas that of the 1000-Hz probe tone was very high (90-100%). In infants older than 13 months, however, the sensitivities of the 226- and 1000-Hz probe tones were 76.2% and 85.7%, respectively. Regarding specificity, the difference between the two probe tones was not significant for any age group. This study confirmed the superiority of the 1000-Hz probe tone for evaluating the middle ear in infants. We recommend using a 1000-Hz probe tone at least up to the age of 12 months for Korean

AP1000 shield building: a constructability challenge

International Nuclear Information System (INIS)

Di Giuseppe, Giovanni; Bonanno, Domenico

2010-01-01

The AP1000 Shield Building, an enhanced structure which surrounds the containment vessel, consists of standard Reinforced Concrete (RC) and composite Steel and Concrete (SC) construction. In the SC module the surface steel plates, (with attached shear studs and angles) filled with concrete, act as the steel reinforcement in concrete. This is a relatively new design technology that required the appropriate use of structural codes, supplemented with information from applicable tests on similar composite steel and concrete construction. Being a newer design concept, existing codes do not provide explicit guidance on SC construction so a review of literature and test data on composite structures similar to AP1000 shield building was done in order to confirm the technical basis for the design. The SC walls, air inlet structure and roof of the Shield Building will be constructed using modular construction practices and then transported to site and lifted into place. These modules, working also as permanent form-work, will be filled with high strength Self- Consolidating Concrete. (SCC) This paper provides a focused and integrated presentation of the enhanced shield building design methodology, testing, constructability and inspection. (authors)

Research on alternative measures in the F1000 system with Google Scholar citation index

OpenAIRE

Saeideh Ebrahimy; Fatemeh Setareh

2016-01-01

The purpose of this study was to assess the relationship between altmetrics measures of F1000 website and citation Indexes of Google scholar. Applied research and survey methods, citation analysis, and webometrics with descriptive analytical approach are used. The study population comprised 218 biomedical research papers have been indexed in F1000 system in the period 2012-2014. A sample of 100 research articles were purposely selected from F1000 system based on the purpose of the research. V...

Domestic design and validation of natural circulation steam generator of China 1000 MWe PWR NPP

International Nuclear Information System (INIS)

Liu, H.Y.; Wang, X.Y.; Wu, G.; Qin, J.M.; Xiong, Ch.H.; Wang, W.; Chen, J.L.; Cheng, H.P.; Zuo, Ch.P.

2005-01-01

In order to meet the requirements of domestic design of China intending built NPP projects, Research Institute of Nuclear Power Operation (RINPO) has achieved design of 1000 MWe NPP steam generator, called RINSG-1000(means 1000MWe SG designed by RINPO), which is based on SG research ,experiments and service experience accumulated by RINPO in more 40 years. Testing validation of two steam generator key technologies, advanced moisture separate device and sludge collector, has been accomplished during the period of 2000 to 2002. This paper describes the design features of RINSG-1000, and provides some validation test results. (authors)

42 CFR 1000.30 - Definitions specific to Medicaid.

Science.gov (United States)

2010-10-01

... 42 Public Health 5 2010-10-01 2010-10-01 false Definitions specific to Medicaid. 1000.30 Section... Medicaid. As used in connection with the Medicaid program, unless the context indicates otherwise— Applicant means an individual whose written application for Medicaid has been submitted to the agency...

Application of computerized procedure system for AP1000

International Nuclear Information System (INIS)

Wen Fang

2012-01-01

With the tendency of digitalisation in instrumentation and control system of nuclear power plants, AP1000, as well as other advanced PWRs, is provided with the technical basis for the application of CPS (computerized procedure system). This paper makes a brief introduction on CPS construction and function. CPS, as an advanced procedure system, does not only have the function of electronic indication for operation procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. In addition, advantages of CPS compared with paper-based procedures and comparison with digital operating system of one M310 plus advanced nuclear power plant are described. Moreover, based on current situation, this paper offers several suggestions on CPS localization for Sanmen AP1000 nuclear power project. Besides, the last part of this paper discusses problems we might meet during the process of CPS localization. (author)

Application of a Russian nuclear reactor simulator VVER-1000

International Nuclear Information System (INIS)

Lopez-Peniche S, A.; Salazar S, E.

2012-10-01

The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

Research on quality assurance classification methodology for domestic AP1000 nuclear power projects

International Nuclear Information System (INIS)

Bai Jinhua; Jiang Huijie; Li Jingyan

2012-01-01

To meet the quality assurance classification requirements of domestic nuclear safety codes and standards, this paper analyzes the quality assurance classification methodology of domestic AP1000 nuclear power projects at present, and proposes the quality assurance classification methodology for subsequent AP1000 nuclear power projects. (authors)

77 FR 56241 - Notice of Withdrawal of Final Design Approval; Westinghouse Electric Company; Advanced Passive 1000

Science.gov (United States)

2012-09-12

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0131] Notice of Withdrawal of Final Design Approval; Westinghouse Electric Company; Advanced Passive 1000 By letter dated December 10, 2010, Westinghouse Electric... final design approval (FDA) for the Advanced Passive 1000 (AP1000) design upon the completion of...

25 CFR 1000.225 - Are meetings or discussions mandatory?

Science.gov (United States)

2010-04-01

... FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Waiver of Regulations § 1000.225 Are meetings or discussions mandatory? No, a meeting with the...

Comparison of 1000-Centistoke versus 5000-Centistoke Silicone Oil in Complex Retinal Detachment Surgery.

Science.gov (United States)

Zafar, Shakir; Shakir, Munira; Mahmood, Syed Asaad; Amin, Saima; Iqbal, Zafar

2016-01-01

To compare the efficacy and complications of using 1000-centistoke versus 5000-centistoke silicone oil for complex retinal detachment repair. Case series. LRBT Tertiary Eye Hospital, Karachi, from January 2007 to June 2013. Eighty-five eyes (85 patients) presenting with superior rhegmatogenous retinal detachments associated with PVR grades B and C (involving not more than 3 clock hours) were randomized to either 1000 centistokes (n=44) or 5000 centistokes (n=41) silicone oil group. All patients underwent 23-gauge pars plana vitrectomy surgery with silicone oil intraocular tamponade. Patient data was analysed at 18 months post-operatively. IBM SPSS 21 was used for data analysis. There were 52 male and 33 female patients aged between 22 and 70 years (45.2 ±16.2). After the first surgery, successful reattachment of the retina was achieved in 67 eyes (78.8%); of which 35 eyes were in 1000-centistoke and 32 eyes in 5000-centistoke groups. Mean pre-operative Best Corrected Visual Acuity (BCVA) was 1.63 ±0.54 which was improved to a mean post-operative BCVAof 1.46 ±0.78 (1.42 ±0.74 in 1000-centistoke group; 1.49 ±0.78 in 5000 centistoke group). The 1000-centistoke group had a significantly higher frequency of oil emulsification which necessitated early removal of silicone oil. There were 66 eyes (77%) with at least one complication (34 eyes in 1000-centistoke group; 32 eyes in 5000-centistoke group) including cataract, corneal abnormalities, raised IOP, hypotony, vitreous haemorrhage and retinal redetachment. Although visual and anatomical outcomes were comparable between the two groups, the 1000-centistoke silicone oil group developed early oil emulsification necessitating its early removal.

Enhanced EGFP Fluorescence Emission in Presence of PEG Aqueous Solutions and PIB1000-PEG6000-PIB1000 Copolymer Vesicles

Directory of Open Access Journals (Sweden)

Noor Muhammad

2013-01-01

Full Text Available An EGFP construct interacting with the PIB1000-PEG6000-PIB1000 vesicles surface reported a ~2-fold fluorescence emission enhancement. Because of the constructs nature with the amphiphilic peptide inserted into the PIB core, EGFP is expected to experience a “pure” PEG environment. To unravel this phenomenon PEG/water solutions at different molecular weights and concentrations were used. Already at ~1 : 10 protein/PEG molar ratio the increase in fluorescence emission is observed reaching a plateau correlating with the PEG molecular weight. Parallel experiments in presence of glycerol aqueous solutions did show a slight fluorescence enhancement however starting at much higher concentrations. Molecular dynamics simulations of EGFP in neat water, glycerol, and PEG aqueous solutions were performed showing that PEG molecules tend to “wrap” the protein creating a microenvironment where the local PEG concentration is higher compared to its bulk concentration. Because the fluorescent emission can be perturbed by the refractive index surrounding the protein, the clustering of PEG molecules induces an enhanced fluorescence emission already at extremely low concentrations. These findings can be important when related to the use of EGFP as reported in molecular biology experiments.

Accuracy of Topcon CM-1000 videokeratoscope on spherical test surfaces.

Science.gov (United States)

Pérez-Yern, E; Fimia-Gil, A; Mateos, F; Carretero, L

1997-01-01

Many videokeratoscopes use mathematical formulas to calculate corneal radii; calculations depend on slope, curvature, coordinate position, or focal properties of the surface. Accuracy of each type of videokeratoscope must be evaluated. A controversy exists about whether axial or tangential methods best provide a precise description of corneal shape; therefore results with the Topcon CM-1000 using both methods were evaluated. Measurements were done on black polymethylmethacrylate (PMMA) spherical calibrated surfaces. Lenses were first aligned and measured and then misaligned in different directions and measured. Results for each position were compared with the zero or alignment position. Accuracy of the CM-1000 was high even under extreme misalignment conditions. Tolerance to misalignment was high (about 300 mm). Misalignment-induced variations in the output results were small (usually less than 0.05 mm). However, important variations (more than 1.00 diopter [D]) were found for the lowest measured radius (6 mm). In some cases, small differences between axial and tangential radii for the same point could be found. With the exception of extremely low radii of curvature, the CM-1000 was accurate for measuring spherical surfaces. Further investigation remains to be done on aspheric surfaces and in clinical practice.

AP1000R pressurised water reactor project in china advances toward completion

International Nuclear Information System (INIS)

Harrop, G.

2014-01-01

The AP1000 R pressurised water reactor (PWR) project in China is the first deployment of its first-of-a-kind Generation III+ technology, making it one of most internationally important and industry-significant new build projects. The innovative AP1000 PWR design contains advanced passive safety and performance features that involve fewer active safety components than a traditional plant, thereby reducing the site footprint. The AP1000 reactor is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission, and it has received an Interim Design Acceptance Confirmation from the Office for Nuclear Regulation and an Interim Statement of Design Acceptability from the Environment Agency in the United Kingdom. Construction and testing of dual AP1000 PWR units is currently in progress in each of two coastal sites in the People's Republic of China: Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Since the initial contract award in 2007, the Westinghouse Consortium has worked in concert with the owners to construct the plants using innovative structural and mechanical modules. Uniquely designed plant components and essential instrumentation and control systems have been manufactured, delivered, and installed at the plants. Numerous personnel, including future reactor operators, have been trained at both the Sanmen and Haiyang sites, and technology transfer of technical documents and computer codes is well underway. The commercial operation dates are now nearing for Sanmen Unit 1 and Haiyang Unit 1, the first two units scheduled for completion. Consequently, these units are now in advanced stages of completion and present activities include planning and preparation for pre-operational testing, system turnover, and commissioning leading to fuel load, and eventual commercial operation. These activities are pioneering, in that they have never before been performed for a new build of

Structural mechanisms of the flux effect for VVER-1000 reactor pressure vessel materials

International Nuclear Information System (INIS)

Gurovich, B.; Kuleshova, E.; Fedotova, S.; Maltsev, D.; Zabusov, O.; Frolov, A.; Erak, D.; Zhurko, D.

2015-01-01

To justify the lifetime extension of VVER-1000 reactor pressure vessels (RPV) up to 60 years and more it is necessary to expand the existing surveillance samples database to beyond design fluence by means of accelerated irradiation in a research reactor. Herewith since the changes in mechanical properties of materials under irradiation are due to occurring structural changes, correct analysis of the data obtained at accelerated irradiation of VVER-1000 RPV materials requires a clear understanding of the structural mechanisms that are responsible for the flux effect in VVER-1000 RPV steels. Two mechanisms are responsible for radiation embrittlement of VVER-1000 RPV steels: the hardening one (radiation hardening due to formation of radiation-induced Ni-based precipitates and radiation defects) and non-hardening one (due to formation of impurities segregations at grain boundaries - reversible temper brittleness). In this context for an adequate interpretation of the mechanical tests results when justifying the lifetime extension of existing units a complex of comparative structural studies (TEM, SEM and AES) of VVER-1000 RPV materials irradiated in different conditions (in research reactor IR-8 and within surveillance samples) was performed. It is shown that the flux effect is observed for materials with high nickel content (weld metals with Ni content > 1.35%) and it is mostly due to the contribution of non-hardening mechanism of radiation embrittlement (the difference in the accumulation kinetics of grain boundary phosphorus segregation) and somewhat contribution of the hardening mechanism (the difference in density of radiation-induced precipitates). Therefore when analyzing the results obtained from the accelerated irradiation of VVER-1000 WM the correction for the flux effect should be made. (authors)

VIPRE modeling of VVER-1000 reactor core for DNB analyses

Energy Technology Data Exchange (ETDEWEB)

Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

1995-09-01

Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping

International Nuclear Information System (INIS)

Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.

2010-01-01

This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction

NCBI nr-aa BLAST: CBRC-DNOV-01-1000 [SEVENS

Lifescience Database Archive (English)

Full Text Available CBRC-DNOV-01-1000 ref|XP_001215926.1| conserved hypothetical protein [Aspergillus terre...us NIH2624] gb|EAU33292.1| conserved hypothetical protein [Aspergillus terreus NIH2624] XP_001215926.1 0.54 32% ...

25 CFR 1000.15 - How many additional Tribes/Consortia may participate in self-governance per year?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false How many additional Tribes/Consortia may participate in self-governance per year? 1000.15 Section 1000.15 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN... Participation in Tribal Self-Governance Eligibility § 1000.15 How many additional Tribes/Consortia may...

24 CFR 1000.40 - Do lead-based paint poisoning prevention requirements apply to affordable housing activities...

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Do lead-based paint poisoning prevention requirements apply to affordable housing activities under NAHASDA? 1000.40 Section 1000.40 Housing... AMERICAN HOUSING ACTIVITIES General § 1000.40 Do lead-based paint poisoning prevention requirements apply...

Analysis of risk management during AP1000 equipment technology transfer and localization

International Nuclear Information System (INIS)

Gao Yongjun; Guan Rui

2009-01-01

This article analyzes the risk factors existing in AP1000 equipment technology transfer and localization process by describing the invitation for bid, tender evaluation and contract negotiation process of the third-generation nuclear power plant technology introduction project of China and discusses the classification, evaluation and analysis methods of risks, and puts forward some referential suggestions for the successful introduction of equipment technology for AP1000 nuclear project. (authors)

The miR-1000-p53 pathway regulates apoptosis and virus infection in shrimp.

Science.gov (United States)

Gong, Yi; Ju, Chenyu; Zhang, Xiaobo

2015-10-01

The p53 protein plays an important role in apoptosis which is involved in the immunity of animals. However, effects of the miRNA-mediated regulation of p53 expression on apoptosis and virus infection are not extensively investigated. To address this issue, the miRNA-mediated p53-dependent apoptotic pathway was explored in this study. The results indicated that p53 could regulate the apoptotic activity of Marsupenaeus japonicas shrimp and influence the infection of white spot syndrome virus (WSSV). The further data presented that miR-1000 could target the 3'-untranslated region (3'UTR) of p53 gene. The results of in vivo experiments showed that the miR-1000 overexpression led to significant decreases of shrimp apoptotic activity and the capacity of WSSV infection, while the miR-1000 silencing resulted in significant increases of apoptotic activity and virus infection, indicating that miR-1000 took great effects on apoptosis and virus infection by targeting p53. Therefore, our study revealed a novel mechanism that the miR-1000-p53 pathway regulated apoptosis and virus infection in shrimp. Copyright © 2015 Elsevier Ltd. All rights reserved.

24 CFR 1000.44 - What prohibitions on the use of debarred, suspended, or ineligible contractors apply?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What prohibitions on the use of debarred, suspended, or ineligible contractors apply? 1000.44 Section 1000.44 Housing and Urban Development... § 1000.44 What prohibitions on the use of debarred, suspended, or ineligible contractors apply? In...

Thorough analysis of unorthodox ABO deletions called by the 1000 Genomes project.

Science.gov (United States)

Möller, M; Hellberg, Å; Olsson, M L

2018-02-01

ABO remains the clinically most important blood group system, but despite earlier extensive research, significant findings are still being made. The vast majority of catalogued ABO null alleles are based on the c.261delG polymorphism. Apart from c.802G>A, other mechanisms for O alleles are rare. While analysing the data set from the 1000 Genomes (1000G) project, we encountered two previously uncharacterized deletions, which needed further exploration. The Erythrogene database, complemented with bioinformatics software, was used to analyse ABO in 2504 individuals from 1000G. DNA samples from selected 1000G donors and African blood donors were examined by allele-specific PCR and Sanger sequencing to characterize predicted deletions. A 5821-bp deletion encompassing exons 5-7 was called in twenty 1000G individuals, predominantly Africans. This allele was confirmed and its exact deletion point defined by bioinformatic analyses and in vitro experiments. A PCR assay was developed, and screening of African samples revealed three donors heterozygous for this deletion, which was thereby phenotypically established as an O allele. Analysis of upstream genetic markers indicated an ancestral origin from ABO*O.01.02. We estimate this deletion as the 3rd most common mechanism behind O alleles. A 24-bp deletion was called in nine individuals and showed greater diversity regarding ethnic distribution and allelic background. It could neither be confirmed by in silico nor in vitro experiments. A previously uncharacterized ABO deletion among Africans was comprehensively mapped and a genotyping strategy devised. The false prediction of another deletion emphasizes the need for cautious interpretation of NGS data and calls for strict validation routines. © 2017 International Society of Blood Transfusion.

Improvement of operational performance and increase of safety of WWER-1000/V-392

International Nuclear Information System (INIS)

Kurakov, Y.A.; Dragunov, Y.G.; Podshibiakin, A.K.; Fil, N.S.; Krushelnitsky, V.N.; Berkovich, V.M.

2001-01-01

The national programme of nuclear power development approved by the Russian Federation Government in 1998 considers the design of WWER-1000/V-392 power unit as a priority project of the new generation NPP with improved operational performances and increased safety. The pilot unit of this design (NVAES-2) is licensed for construction at the Novovoronezh NPP site. The NVAES-2 design is developed on the basis of standard power unit with reactor plant V-320. Twenty units of this type are in operation at the nuclear power plants in Russia, Ukraine and Bulgaria having totally about 270 reactor-years of operation. Two more V-320 units are being commissioned this year at Rostov NPP and Temelin NPP. So, the WWER-1000/V-392 design is as a whole an evolutionary development of the operating standard unit WWER-1000/V-320. Many technical solutions aimed at increase of safety and improvement of operational performance of the plant are implemented in the NVAES-2 design, such as advanced reactor WWER-1000, passive system of residual power removal, passive system of the core flooding under loss-of-coolant accidents, and others. NVAES-2 design refers to a class of advanced light water reactors and corresponds to the international requirements imposed to the nuclear power plants to be put into operation after the year 2000. New V-392 power unit has a good perspective from the view point of extensive implementation in the framework of the nuclear electricity production in Russia. Design decisions on NVAES-2 power unit with WWER-1000/V-392 reactor plant which assure significantly higher safety level and improve economical performance as compared to the operating WWER-1000 units are briefly considered in the present paper. (author)

25 CFR 1000.302 - “What is imminent jeopardy” to a trust asset?

Science.gov (United States)

2010-04-01

.../or Federal law while performing trust functions if the disregard creates such an immediate threat. ... 25 Indians 2 2010-04-01 2010-04-01 false âWhat is imminent jeopardyâ to a trust asset? 1000.302...-DETERMINATION AND EDUCATION ACT Reassumption § 1000.302 “What is imminent jeopardy” to a trust asset? Imminent...

25 CFR 1000.175 - What is the process for conducting the negotiation phase?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What is the process for conducting the negotiation phase? 1000.175 Section 1000.175 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Negotiation...

Carbon Nanostructure of Kraft Lignin Thermally Treated at 500 to 1000 °C.

Science.gov (United States)

Zhang, Xuefeng; Yan, Qiangu; Leng, Weiqi; Li, Jinghao; Zhang, Jilei; Cai, Zhiyong; Hassan, El Barbary

2017-08-21

Kraft lignin (KL) was thermally treated at 500 to 1000 °C in an inert atmosphere. Carbon nanostructure parameters of thermally treated KL in terms of amorphous carbon fraction, aromaticity, and carbon nanocrystallites lateral size ( L a ), thickness ( L c ), and interlayer space ( d 002 ) were analyzed quantitatively using X-ray diffraction, Raman spectroscopy, and high-resolution transmission electron microscopy. Experimental results indicated that increasing temperature reduced amorphous carbon but increased aromaticity in thermally treated KL materials. The L c value of thermally treated KL materials averaged 0.85 nm and did not change with temperature. The d 002 value decreased from 3.56 Å at 500 °C to 3.49 Å at 1000 °C. The L a value increased from 0.7 to 1.4 nm as temperature increased from 500 to 1000 °C. A nanostructure model was proposed to describe thermally treated KL under 1000 °C. The thermal stability of heat treated KL increased with temperature rising from 500 to 800 °C.

The Belgian initiative G1000. A feasible model of deliberative cyber democracy

Directory of Open Access Journals (Sweden)

Laura Tejedor Fuentes

2016-11-01

Full Text Available In 2010, after five hundred days without any stable form of government, a group of Belgian citizens from a wide range of social backgrounds launched a citizenship summit called 'G1000'. The aim of the event was to create citizen involvement in politics by stimulating participants to take part in debates about social issues and by applying innovative techniques of deliberative democracy. In order to verify whether the model mentioned meets the criteria of an “online democratic-deliberative” experiment, this article carried out a quantitative analysis of the G1000 based on the work of Fishkin, Chadwick and Habermas. According to the majority of authors who have contributed to the topic, including the spokesperson and organizer of the event, Vincent Jacquet the results demonstrate that the G1000 summit is, in fact, a genuine democratic-deliberative initiative.

Genome update: the 1000th genome - a cautionary tale

DEFF Research Database (Denmark)

Lagesen, Karin; Ussery, David; Wassenaar, Gertrude Maria

2010-01-01

conclusions for example about the largest bacterial genome sequenced. Biological diversity is far greater than many have thought. For example, analysis of multiple Escherichia coli genomes has led to an estimate of around 45 000 gene families more genes than are recognized in the human genome. Moreover......There are now more than 1000 sequenced prokaryotic genomes deposited in public databases and available for analysis. Currently, although the sequence databases GenBank, DNA Database of Japan and EMBL are synchronized continually, there are slight differences in content at the genomes level...... for a variety of logistical reasons, including differences in format and loading errors, such as those caused by file transfer protocol interruptions. This means that the 1000th genome will be different in the various databases. Some of the data on the highly accessed web pages are inaccurate, leading to false...

Operational indices of WWER-1000 fuel assemblies and their improvements

Energy Technology Data Exchange (ETDEWEB)

Vasilchenko, I; Demin, E [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

1994-12-31

The most general design features of WWER-1000 fuel assembly are discussed. The following advantages of design are stated as well as their operational confirmation and occurrences: (1) `packing` density (tight-lattice) of fuel rods within the fuel assemblies; (2) simple handling of fuel assemblies and its small vulnerability; (3) good conditions for coolant mixing; (4) protection of the absorber rods against coolant effect; (5) adaptability to manufacture that provides stable quality. The main operational indices gathered during a ten-year period (1982-1992) at 17 WWER-1000 units in Russia and Ukraine are outlined. Provisions for emergency protection reliability are described. Future directions to improve fuel economy and control rod operability are discussed. 1 fig.

Operational indices of WWER-1000 fuel assemblies and their improvements

International Nuclear Information System (INIS)

Vasilchenko, I.; Demin, E.

1994-01-01

The most general design features of WWER-1000 fuel assembly are discussed. The following advantages of design are stated as well as their operational confirmation and occurrences: 1) 'packing' density (tight-lattice) of fuel rods within the fuel assemblies; 2) simple handling of fuel assemblies and its small vulnerability; 3) good conditions for coolant mixing; 4) protection of the absorber rods against coolant effect; 5) adaptability to manufacture that provides stable quality. The main operational indices gathered during a ten-year period (1982-1992) at 17 WWER-1000 units in Russia and Ukraine are outlined. Provisions for emergency protection reliability are described. Future directions to improve fuel economy and control rod operability are discussed. 1 fig

WWER-1000 fuel cycles: current situation and outlook

International Nuclear Information System (INIS)

Kosourov, E.; Pavlov, V.; Pavlovichev, A.; Spirkin, E.; Shcherenko, A.

2013-01-01

Usage mode of nuclear fuel in WWER type reactor has been changed significantly till the moment of the first WWER-1000 commissioning. There are a lot of improvements, having an impact on the fuel cycle, have been implemented for units with WWER-1000. FA design and its constructional materials, FA fuel weight, burnable poison, usage mode of units and etc have been modified. As the result of development it has been designed a modern FA with rigid skeleton. As a whole it allows to use more efficient configurations of the core, to extend range of fuel cycle lengths and to provide good flexibility in the operation. In recent years there were in progress works on increasing FA uranium capacity. As the result there were developed two designs of the fuel rod: 1) the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively and 2) the fuel column height of 3530 mm, the fuel pellet diameter of 7.8 mm without the central hole. Such fuel rods have operating experience as a part of different FA designs. Positive operating experience was a base of new FA (TVS-4) development with the fuel column height of 3680 mm and the fuel pellet diameter of 7.8 mm without the central hole. The paper presents the overview of WWER-1000, AES-2006 and WWER-TOI fuel cycles based on FAs with fuel rod designs described above. There are demonstrated fuel cycle possibilities and its technical and economic characteristics. There are discussed problems of further fuel cycle improvements (fuel enrichment increase above 5 %, use of erbium as alternative burnable poison) and their impact on neutronics characteristics. (authors)

Safety of NPP with WWER-440 and WWER-1000 reactors

Energy Technology Data Exchange (ETDEWEB)

Balabanov, E [Energoproekt, Sofia (Bulgaria); Gledachev, J; Angelov, D [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1996-12-31

The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab.

Safety of NPP with WWER-440 and WWER-1000 reactors

International Nuclear Information System (INIS)

Balabanov, E.; Gledachev, J.; Angelov, D.

1995-01-01

The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab

Reevaluation of the PMS alarm set-points in OPR1000

International Nuclear Information System (INIS)

Roh, Kyung Ho; Yang, Sung Tae; Jung, Sung In

2011-01-01

In Optimized Power Reactor 1000 (OPR1000), the common alarm of the plant monitoring system (PMS), which is related to the channel-to-channel core protection calculator (CPC), experiences frequent deviations in the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) between the middle of the cycle and the end of the cycle. Because the channel-to-channel CPC causes deviations in the values of the DNBR and LPD, the increase in the CPC input variables exceeds the alarm set-points. The CPC DNBR and LPD are defined as follows: Deviation = the average of four CPC channels of the LPD and DNBR minus the value of each CPC channel LPD and DNBR. In this paper, we report on a review by the Korea Hydro and Nuclear Power Co., Ltd. (KHNP) regarding the suitability of the alarm set-points for the channel-tochannel deviations of the CPC DNBR and LPD. The set-points were revaluated in light of operational experience and the case of Palo Verde (which is the reference model of OPR1000). The KHNP consequently revised the relevant procedures, as well as and the PMS alarm set-points, as part of its follow-up action

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R+D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR 1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

Thermal aging effects of VVER-1000 weld metal under operation temperature

International Nuclear Information System (INIS)

Chernobaeva, A.A.; Kuleshova, E.A.; Gurovich, B.A.; Erak, D.Y.; Zabusov, O.O.; Maltsev, D.A.; Zhurko, D.A.; Papina, V.B.; Skundin, M.A.

2015-01-01

The VVER-1000 thermal aging surveillance specimen sets are located in the reactor pressure vessel (RPV) under real operation conditions. Thermal aging surveillance specimens data are the most reliable source of the information about changing of VVER-1000 RPV materials properties because of long-term (hundred thousand hours) exposure at operation temperature. A revision of database of VVER-1000 weld metal thermal aging surveillance specimens has been done. The reassessment of transition temperature (T t ) for all tested groups of specimens has been performed. The duration of thermal exposure and phosphorus contents have been defined more precisely. The analysis of thermal aging effects has been done. The yield strength data, study of carbides evolution show absence of hardening effects due to thermal aging under 310-320 C degrees. Measurements of phosphorus content in grain boundaries segregation in different states have been performed. The correlation between intergranular fracture mode in Charpy specimens and transition temperature shift under thermal aging at temperature 310-320 C degrees has been revealed. All these data allow developing the model of thermal aging. (authors)

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Meseth, J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R and D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

WWER-1000 nuclear fuel manufacturing process at PJSC MSZ

International Nuclear Information System (INIS)

Morylev, A.; Bagdatyeva, E.; Aksenov, P.

2015-01-01

In this report a brief description of WWER-1000 fuel manufacturing process steps at PJSC MSZ as: uranium dioxide powder fabrication; fuel pellet manufacture fuel rod manufacture working assembly and fuel assembly manufacture is given. The implemented innovations are also presented

Pacing Behavior and Tactical Positioning in 500- and 1000-m Short-Track Speed Skating.

Science.gov (United States)

Noorbergen, Olaf S; Konings, Marco J; Micklewright, Dominic; Elferink-Gemser, Marije T; Hettinga, Florentina J

2016-09-01

To explore pacing behavior and tactical positioning during the shorter 500- and 1000-m short-track competitions. Lap times and intermediate rankings of elite 500- and 1000-m short-track-skating competitors were collected over the 2012-13 season. First, lap times were analyzed using a MANOVA, and for each lap, differences between sex, race type, final ranking, and stage of competition were determined. Second, Kendall tau-b correlations were used to assess relationships between intermediate and final rankings. In addition, intermediate rankings of the winner of each race were examined. Top-placed athletes appeared faster than bottom-placed athletes in every lap in the 500-m, while in the 1000-m no differences were found until the final 4 laps (P < .05). Correlations between intermediate and final rankings were already high at the beginning stages of the 50-m (lap 1: r = .59) but not for the 1000-m (lap 1: r = .21). Although 500- and 1000-m short-track races are both relatively short, fundamental differences in pacing behavior and tactical positioning were found. A fast-start strategy seems to be optimal for 500-m races, while the crucial segment in 1000-m races seems to be from the 6th lap to the finish line (ie, after ± 650 m). These findings provide evidence to suggest that athletes balance between choosing an energetically optimal profile and the tactical and positional benefits that play a role when riding against an opponent, as well as contributing to developing novel insights in exploring athletic behavior when racing against opponents.

Alignment of 1000 Genomes Project reads to reference assembly GRCh38.

Science.gov (United States)

Zheng-Bradley, Xiangqun; Streeter, Ian; Fairley, Susan; Richardson, David; Clarke, Laura; Flicek, Paul

2017-07-01

The 1000 Genomes Project produced more than 100 trillion basepairs of short read sequence from more than 2600 samples in 26 populations over a period of five years. In its final phase, the project released over 85 million genotyped and phased variants on human reference genome assembly GRCh37. An updated reference assembly, GRCh38, was released in late 2013, but there was insufficient time for the final phase of the project analysis to change to the new assembly. Although it is possible to lift the coordinates of the 1000 Genomes Project variants to the new assembly, this is a potentially error-prone process as coordinate remapping is most appropriate only for non-repetitive regions of the genome and those that did not see significant change between the two assemblies. It will also miss variants in any region that was newly added to GRCh38. Thus, to produce the highest quality variants and genotypes on GRCh38, the best strategy is to realign the reads and recall the variants based on the new alignment. As the first step of variant calling for the 1000 Genomes Project data, we have finished remapping all of the 1000 Genomes sequence reads to GRCh38 with alternative scaffold-aware BWA-MEM. The resulting alignments are available as CRAM, a reference-based sequence compression format. The data have been released on our FTP site and are also available from European Nucleotide Archive to facilitate researchers discovering variants on the primary sequences and alternative contigs of GRCh38. © The Authors 2017. Published by Oxford University Press.

Accident Analysis of Chinese CPR1000 in Response to Station Blackout

Energy Technology Data Exchange (ETDEWEB)

Kim, Juyoul [FNC Technology Co., Yongin (Korea, Republic of); Cilliers, Anthonie [North-West University, Potchefstroom (South Africa)

2016-10-15

Stress tests required evaluation of the consequences of loss of safety functions from any initiating event (e.g., earthquake or flooding) causing loss of electrical power, including station blackout (SBO). The SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel generators, failure of alternate current (AC) power restoration and the eventual degradation of the reactor coolant pump (RCP) seals resulting in a long term loss of coolant. Using PCTRAN/CPR1000, this study analyses the station blackout on a Chinese CPR1000 which is the most representative type reactor in terms of number of reactors, operating period, power capacity and geological distance from Korean Peninsula. Both the physical effects of the accidents as well as the releases of radioisotopes are calculated and discussed. Station blackout simulation was conducted in this study. The resulting effects seen are consistent with other stress test station blackout tests used utilizing licensed simulation codes. An exact comparison is however not possible as the plants on which the simulations was done vary greatly and the limitations of availability to Chinese FSAR. PCTRAN/CPR1000 is an extremely useful simulation package that provides engineers and scientists very accurate feedback to how a nuclear power plant would react as a whole under various plant conditions. It is able to do this extremely fast as well. As a training tool PCTRAN/CPR1000 provides hands-on experience with many of the primary plant operations and develops an intuitive understanding of the plant.

Scalable synthesis and post-modification of a mesoporous metal-organic framework called NU-1000.

Science.gov (United States)

Wang, Timothy C; Vermeulen, Nicolaas A; Kim, In Soo; Martinson, Alex B F; Stoddart, J Fraser; Hupp, Joseph T; Farha, Omar K

2016-01-01

The synthesis of NU-1000, a highly robust mesoporous (containing pores >2 nm) metal-organic framework (MOF), can be conducted efficiently on a multigram scale from inexpensive starting materials. Tetrabromopyrene and (4-(ethoxycarbonyl)phenyl)boronic acid can easily be coupled to prepare the requisite organic strut with four metal-binding sites in the form of four carboxylic acids, while zirconyl chloride octahydrate is used as a precursor for the well-defined metal oxide clusters. NU-1000 has been reported as an excellent candidate for the separation of gases, and it is a versatile scaffold for heterogeneous catalysis. In particular, it is ideal for the catalytic deactivation of nerve agents, and it shows great promise as a new generic platform for a wide range of applications. Multiple post-synthetic modification protocols have been developed using NU-1000 as the parent material, making it a potentially useful scaffold for several catalytic applications. The procedure for the preparation of NU-1000 can be scaled up reliably, and it is suitable for the production of 50 g of the tetracarboxylic acid containing organic linker and 200 mg-2.5 g of NU-1000. The entire synthesis is performed without purification by column chromatography and can be completed within 10 d.

Role of a Modulator in the Synthesis of Phase-Pure NU-1000.

Science.gov (United States)

Webber, Thomas E; Liu, Wei-Guang; Desai, Sai Puneet; Lu, Connie C; Truhlar, Donald G; Penn, R Lee

2017-11-15

NU-1000 is a robust, mesoporous metal-organic framework (MOF) with hexazirconium nodes ([Zr 6 O 16 H 16 ] 8+ , referred to as oxo-Zr 6 nodes) that can be synthesized by combining a solution of ZrOCl 2 ·8H 2 O and a benzoic acid modulator in N,N-dimethylformamide with a solution of linker (1,3,6,8-tetrakis(p-benzoic acid)pyrene, referred to as H 4 TBAPy) and by aging at an elevated temperature. Typically, the resulting crystals are primarily composed of NU-1000 domains that crystallize with a more dense phase that shares structural similarity with NU-901, which is an MOF composed of the same linker molecules and nodes. Density differences between the two polymorphs arise from the differences in the node orientation: in NU-1000, the oxo-Zr 6 nodes rotate 120° from node to node, whereas in NU-901, all nodes are aligned in parallel. Considering this structural difference leads to the hypothesis that changing the modulator from benzoic acid to a larger and more rigid biphenyl-4-carboxylic acid might lead to a stronger steric interaction between the modulator coordinating on the oxo-Zr 6 node and misaligned nodes or linkers in the large pore and inhibit the growth of the more dense NU-901-like material, resulting in phase-pure NU-1000. Side-by-side reactions comparing the products of synthesis using benzoic acid or biphenyl-4-carboxylic acid as a modulator produce structurally heterogeneous crystals and phase-pure NU-1000 crystals. It can be concluded that the larger and more rigid biphenyl-4-carboxylate inhibits the incorporation of nodes with an alignment parallel to the neighboring nodes already residing in the crystal.

P sub(T)-flow in 100 - 1000 TeV

International Nuclear Information System (INIS)

Amato, N.M.; Arata, N.; Maldonado, R.H.C.

1983-01-01

It is studied how the flow of transverse momenta (Σ P sub(T) or Σ ER) behaves in multi-particle production phenomena in the region of visible energy 10 -1000 TeV, using the emulsion chamber data of Brazil-Japan Collaboration. (Author) [pt

Results of operation of VVER-1000 FAs manufactured at PJSC NCCP

International Nuclear Information System (INIS)

Davidov, D.; Brovkin, O.; Bezborodov, Y.

2015-01-01

Fuel Assemblies manufactured at PJSC NCCP are in operation at 27 VVER-1000 power units at 11 NPPs in Russia, Ukraine, Bulgaria, China, Iran and India. Basic results of operation of PJSC NCCP VVER-1000 FAs during 2007-2014 are presented. The operation results confirm the design characteristics of fuel, i.e.: average fuel burnup up to 55 MW*day/kgU in FAs; safe and reliable FA operation, with low leaking rate (in the order of 10-6). The achieved operation characteristics of TVSA and TVS-2M Fuel Assemblies prove the quality, reliability and competitiveness of FAs manufactured at PJSC NCCP

A calculation method for transient flow distribution of SCWR(CSR1000)

International Nuclear Information System (INIS)

Chen, Juan; Zhou, Tao; Chen, Jie; Liu, Liang; Muhammad, Ali Shahzad; Muhammad, Zeeshan Ali; Xia, Bangyang

2017-01-01

The supercritical water reactor CSR1000 is selected for the study. A parallel channel flow transient flow distribution module is developed, which is used for solving unsteady nonlinear equations. The incorporated programs of SCAC-CSR1000 are executed on normal and abnormal operating conditions. The analysis shows that: 1. Transient flow distribution can incorporate parallel channel flow calculation, with an error less than 0.1%; 2. After a total loss of coolant flow, the flow of each channel shows a downward trend; 3. In the event of introducing a traffic accident, the first coolant flow shows an increasing trend.

Production Facility Prototype Blower 1000 Hour Test Results II

Energy Technology Data Exchange (ETDEWEB)

Wass, Alexander Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Woloshun, Keith Albert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dalmas, Dale Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Frank Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-08

Long duration tests of the Aerzen GM 12.4 roots style blower in a closed loop configuration provides valuable data and lessons learned for long-term operation at the Mo-99 production facility. The blower was operated in a closed loop configuration with the flow conditions anticipated in plant operation with a Mo-100 target inline. The additional thermal energy generated from beam heating of the Mo-100 disks were not included in these tests. Five 1000 hour tests have been completed since the first test was performed in January of 2016. All five 1000 hour tests have proven successful in exposing preventable issues related to oil and helium leaks. All blower tests to this date have resulted in stable blower performance and consistency. A summary of the results for each test, including a review of the first and second tests, are included in this report.

38 CFR 17.1000 - Payment or reimbursement for emergency services for nonservice-connected conditions in non-VA...

Science.gov (United States)

2010-07-01

... for emergency services for nonservice-connected conditions in non-VA facilities. 17.1000 Section 17.1000 Pensions, Bonuses, and Veterans' Relief DEPARTMENT OF VETERANS AFFAIRS MEDICAL Payment Or Reimbursement for Emergency Services for Nonservice-Connected Conditions in Non-Va Facilities § 17.1000 Payment...

Pacing Behavior and Tactical Positioning in 500-and 1000-m Short-Track Speed Skating

NARCIS (Netherlands)

Noorbergen, Olaf S.; Konings, Marco J.; Micklewright, Dominic; Elferink-Gemser, Marge T.; Hettinga, Florentina J.

Purpose: To explore pacing behavior and tactical positioning during the shorter 500- and 1000-m short-track competitions. Methods: Lap times and intermediate rankings of elite 500- and 1000-m short-track-skating competitors were collected over the 2012-13 season. First, lap times were analyzed using

24 CFR 1000.24 - If an Indian tribe assumes environmental review responsibility, how will HUD assist the Indian...

Science.gov (United States)

2010-04-01

...? 1000.24 Section 1000.24 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES General § 1000.24 If an Indian tribe assumes...

Liquid radioactive waste processing system in Improved OPR-1000

International Nuclear Information System (INIS)

Lee, Soonmin; Kim, Kiljung; Park, Jungsu

2008-01-01

The design goal of liquid rad waste system is to minimize the release of radioactive materials to the environment, the occupational radiation exposure to workers, and the solid rad waste volume generated from LRS operation. In 1998, KOPEC in conjunction with KHNP (Korea Hydro and Nuclear Power Co.) started a special task study which had been focused on the worldwide advanced technologies in the liquid rad waste process area by considering the design goals above. As a result of this task, KOPEC and KHNP finally decided to adopt a reverse osmosis processing method for Improved OPR-1000 in Korea. The advanced LRS design incorporating the R/O process has been introduced into Shin-Wolsong 1 and 2 (SWN 1 and 2) as well as Shin-Kori 1 and 2 (SKN 1 and 2), which are recently under construction, and also is adopted for Shin-Kori 3 and 4 (SKN 3 and 4) and Shin-Ulchin 1 and 2 (SUN 1 and 2), which are planned for the near future construction as the first APR-1400 type of Korean reactors. The LRS shop performance test for SKN 1 and 2 (Improved OPR-1000 R/O package system) was conducted by DOOSAN and DTS (Diversified Technologies Services, Inc) in January, 2008. The purpose of the test was to demonstrate the performance of actual R/O system to be installed in SKN 1 and 2 site. In this paper, overall system configuration and the shop performance test result is presented based on Improved OPR-1000 LRS R/O Package system

Comparison of HapMap and 1000 Genomes Reference Panels in a Large-Scale Genome-Wide Association Study

DEFF Research Database (Denmark)

de Vries, Paul S; Sabater-Lleal, Maria; Chasman, Daniel I

2017-01-01

An increasing number of genome-wide association (GWA) studies are now using the higher resolution 1000 Genomes Project reference panel (1000G) for imputation, with the expectation that 1000G imputation will lead to the discovery of additional associated loci when compared to HapMap imputation. In...

25 CFR 1000.224 - How can a Tribe/Consortium expedite the review of a regulation waiver request?

Science.gov (United States)

2010-04-01

... regulation waiver request? 1000.224 Section 1000.224 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Waiver of Regulations § 1000.224 How can a...

24 CFR 1000.242 - When does the requirement for exemption from taxation apply to affordable housing activities?

Science.gov (United States)

2010-04-01

... exemption from taxation apply to affordable housing activities? 1000.242 Section 1000.242 Housing and Urban... ACTIVITIES Indian Housing Plan (IHP) § 1000.242 When does the requirement for exemption from taxation apply to affordable housing activities? The requirement for exemption from taxation applies only to rental...

Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-07

At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

International Nuclear Information System (INIS)

1994-01-01

At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

Heterogeneous LTE-Advanced Network Expansion for 1000x Capacity

DEFF Research Database (Denmark)

Hu, Liang; Sanchez, Maria Laura Luque; Maternia, Michal

2013-01-01

this paper studies LTE (Long-Term Evolution)-Advanced heterogeneous network expansion in a dense urban environment for a 1000 times capacity increase and a 10 times increase in minimum user data rate requirements. The radio network capacity enhancement via outdoor and indoor small cell densificat...

7 CFR 1000.42 - Classification of transfers and diversions.

Science.gov (United States)

2010-01-01

... MARKETING SERVICE (Marketing Agreements and Orders; Milk), DEPARTMENT OF AGRICULTURE GENERAL PROVISIONS OF FEDERAL MILK MARKETING ORDERS Classification of Milk § 1000.42 Classification of transfers and diversions... continuing in that fashion until the excess diverted milk has been assigned as producer milk under the...

Experience on KKNPP VVER 1000 MWe water chemistry

International Nuclear Information System (INIS)

Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Pillai, Suresh Kumar

2015-01-01

Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration with Russian Federation at Kudankulam in Tirunelveli District, Tamilnadu. Unit - 1 attained criticality on July 13 th 2013 and the unit was synchronized to grid on 22 nd October 2013. This paper highlights experience gained on water chemistry regime for primary and secondary circuit. (author)

25 CFR 1000.136 - When will this determination be made?

Science.gov (United States)

2010-04-01

... EDUCATION ACT Non-BIA Annual Self-Governance Compacts and Funding Agreements Eligibility § 1000.136 When will this determination be made? It will occur during the pre-negotiation process, subject to the...

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

Energy Technology Data Exchange (ETDEWEB)

Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

25 CFR 1000.278 - Does this coverage extend to subcontractors of self-governance AFAs?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false Does this coverage extend to subcontractors of self-governance AFAs? 1000.278 Section 1000.278 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS... subcontractors of self-governance AFAs? No, subcontractors or subgrantees providing services to a Pub. L. 93-638...

7 CFR 1000.53 - Announcement of class prices, component prices, and advanced pricing factors.

Science.gov (United States)

2010-01-01

... advanced pricing factors. 1000.53 Section 1000.53 Agriculture Regulations of the Department of Agriculture..., component prices, and advanced pricing factors. (a) On or before the 5th day of the month, the market... administrator for each Federal milk marketing order shall announce the following prices and pricing factors for...

25 CFR 1000.367 - Will the Department evaluate a Tribe's/Consortium's performance of non-trust related programs?

Science.gov (United States)

2010-04-01

... Evaluations § 1000.367 Will the Department evaluate a Tribe's/Consortium's performance of non-trust related... 25 Indians 2 2010-04-01 2010-04-01 false Will the Department evaluate a Tribe's/Consortium's performance of non-trust related programs? 1000.367 Section 1000.367 Indians OFFICE OF THE ASSISTANT SECRETARY...

25 CFR 1000.352 - What are “trust resources” for the purposes of the trust evaluation process?

Science.gov (United States)

2010-04-01

..., property, assets, or claims, and any intangible right or interest in any of the foregoing; (2) Any other... evaluation process? 1000.352 Section 1000.352 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS... INDIAN SELF-DETERMINATION AND EDUCATION ACT Trust Evaluation Review § 1000.352 What are “trust resources...

24 CFR 1000.110 - Under what conditions may non low-income Indian families participate in the program?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Under what conditions may non low-income Indian families participate in the program? 1000.110 Section 1000.110 Housing and Urban... ACTIVITIES Affordable Housing Activities § 1000.110 Under what conditions may non low-income Indian families...

Comparing genetic variants detected in the 1000 genomes project ...

Indian Academy of Sciences (India)

Single-nucleotide polymorphisms (SNPs) determined based on SNP arrays from the international HapMap consortium (HapMap) and the genetic variants detected in the 1000 genomes project (1KGP) can serve as two references for genomewide association studies (GWAS). We conducted comparative analyses to provide ...

25 CFR 1000.21 - When does a Tribe/Consortium have a “material audit exception”?

Science.gov (United States)

2010-04-01

...-Governance Eligibility § 1000.21 When does a Tribe/Consortium have a “material audit exception”? A Tribe/Consortium has a material audit exception if any of the audits that it submitted under § 1000.17(c...

The Design of Cooling System Model on The AP1000 Containment

International Nuclear Information System (INIS)

Daddy Setyawan; Yerri Noer Kartiko; Aryadi Suwono; Ari Darmawan Pasek; Nathanael P Tandian; Efrizon Umar

2009-01-01

The policy of national energy leads to the utilization of new energy as nuclear energy, and also contains some efforts to increase reactor safety and optimizing in the design of safety system component such as passive cooling system on reactor containment tank. Because of this, the assessment of safety level to passive safety system needs to be made. To increase the understanding it, the design of cooling system model on containment tank should be done to get safety level on cooling system in the AP1000 containment. To reach the similar model with reality and inexpensive cost, we should make assessment about similarity and dimensionless number. While the heat transfer of air natural circulation and water spray cooling system are a result of gravity approach, we can calculate Grashof modification number and Reynolds number respectively. By this approach, we have a factor of forty for laboratory model. From this model, we hope that we get characteristic correlation to heat transfer on the containment of AP1000 for both air natural circulation and water spray result from gravity. Finally, we can assess the safety level of passive cooling system on the AP1000 containment. (author)

Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

International Nuclear Information System (INIS)

Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

2005-01-01

Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

A Neutronic Feasibility Study of an OPR-1000 Core Design with Boron-bearing Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Hoon; Park, Sang Yoon; Lee, Chung Chan; Yang, Yong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

In Westinghouse plants, boron is mainly used as a form of the integral fuel burnable absorber (IFBA) with a thin coating of zirconium diboride (ZrB{sub 2}) or wet annular burnable absorber (WABA) with a hollow Al{sub 2}O{sub 3}+B{sub 4}C pellet. In OPR-1000, on the other hand, gadolinia is currently employed as a form of an admixture which consists of Gd{sub 2}O{sub 3} of 6∼8 w/o and UO{sub 2} of natural uranium. Recently, boron-bearing UO{sub 2} fuel (BBF) with the high density of greater than 94%TD has been developed by using a low temperature sintering technique. In this paper, the feasibility of replacing conventional gadolinia-bearing UO{sub 2} fuel (GBF) in OPR-1000 with newly developed boron-bearing fuel is evaluated. Neutronic feasibility study to utilize the BBF in OPR-1000 core has been performed. The results show that the OPR-1000 core design with the BBF is feasible and promising in neutronic aspects. Therefore, the use of the BBF in OPR-1000 can reduce the dependency on the rare material such as gadolinium. However, the burnout of the {sup 10}B isotope results in helium gas, so fuel performance related study with respect to helium generation is needed.

Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

2009-06-15

ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

Stresses in transition region of VVER-1000 reactor vessels

International Nuclear Information System (INIS)

Namgung, I.; Nguye, T.L.

2014-01-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Stresses in transition region of VVER-1000 reactor vessels

Energy Technology Data Exchange (ETDEWEB)

Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

2014-07-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Results of the 1000 Hour Rotary Microfilter Endurance Test

International Nuclear Information System (INIS)

Herman, D.

2010-01-01

During operation, a primary concern is lifetime of the mechanical seal. Leakage from main shaft mechanical seal is expected to be first indication of wear on rotary filter but will not be the ultimate failure of the filter. Initial clearance inadequate to support thermal expansion resulted in initial scarring of journal. Tolerances adjusted and no journal issues were found during 1000 hour test. Disassembly and inspection of the rotary joint seals showed no unusual wear. No leakage observed during test. Total operation time over 1500 hours. Test summary successfully demonstrated: (1) Filtration of a 'challenging' SRS simulant up to 15 wt % insoluble solids in a 5.6 M salt simulant; (2) Sludge washing; (3) In-situ acid cleaning with dilute acid; (4) Over 1000 hours of operation on new journal material; and (5) Over 1500 hours of operation on all seals. The filter out lasted: 2 air compressors, 2 power outages (one planned), 2 chillers, 1 fire, 1 electrical breaker and 1 feed pump seal.

Assessment of the recovery annealing efficiency for VVER-1000 materials' structure reset and lifetime extension

International Nuclear Information System (INIS)

Gurovich, B.; Kuleshova, E.; Prikhodko, K.; Fedotova, S.

2011-01-01

The results of the VVER-1000 reactor pressure vessels welds studies based on the surveillance specimens sets have revealed a high embrittlement rate of steel with high nickel content compared with predicted embrittlement determined from the Russian Guide. For these critical vessels further safe operation (even during design service life) is not allowed without additional measures (recovery annealing of the VVER-1000 welds as earlier for VVER- 440). The reason is that the rate of high nickel VVER-1000 welds embrittlement is significantly higher than that is for base metal. In order to solve a problem of VVER-1000 lifetime extension recovery annealing validation and accelerated reirradiation of specimens for prolonged operation period estimation after annealing were necessary. In this work comparison of electron-microscopy fine structure studies and fractographic studies of Charpy specimens fracture surface of the VVER-1000 high nickel welds in different states were carried out. It allows estimation of the recovery annealing effect on steels structure and its behavior at further operation. It is shown that both secondary and primary irradiation causes alike radiation-induced fine structure changes: dislocation loops and nano-size precipitates. Recovery annealing leads to full dislocation loops dissolution and significant nano-size precipitates solution but not to the initial values. The rate of radiation defects and radiation-induced precipitates accumulation at reirradiation weld after recovery annealing is lower than at primary irradiation and determine the lower secondary embrittlement rate of VVER-1000 weld. (authors)

PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-04-01

Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

IRMM-1000a and IRMM-1000b. Uranium reference materials certified for the production date based on the 230Th/234U radiochronometer. Part II. Certification

International Nuclear Information System (INIS)

Venchiarutti, C.; Richter, S.; Jakopic, R.; Aregbe, Y.; Varga, Z.; Nicholl, A.; Krajko, J.; Mayer, K.

2016-01-01

The IRMM-1000a and IRMM-1000b uranium reference materials, of 20 and 50 mg uranium, respectively, were produced by the European Commission Joint Research Centre's Institute for Reference Materials and Measurements (EC-JRC-IRMM) in collaboration with the Institute for Transuranium Elements (EC-JRC-ITU). They are novel uranium reference materials certified for the production date based on the 230 Th/ 234 U radiochronometer, i.e. the date of the last chemical separation of these two radionuclides. The certified reference value and its uncertainty, homogeneity and stability of the material were established in accordance with the ISO Guide 34:2009 and the 'Guide to the Expression of Uncertainty in Measurement'. (author)

24 CFR 1000.62 - What is considered program income and what restrictions are there on its use?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What is considered program income and what restrictions are there on its use? 1000.62 Section 1000.62 Housing and Urban Development... § 1000.62 What is considered program income and what restrictions are there on its use? (a) Program...

40 CFR 721.1000 - Benzenamine, 3-chloro-2,6-dinitro-N,N-dipropyl-4-(tri-fluoro-methyl)-.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Benzenamine, 3-chloro-2,6-dinitro-N,N-dipropyl-4-(tri-fluoro-methyl)-. 721.1000 Section 721.1000 Protection of Environment ENVIRONMENTAL... Significant New Uses for Specific Chemical Substances § 721.1000 Benzenamine, 3-chloro-2,6-dinitro-N,N...

Probabilistic safety assessment for Balakovo 1000 MW NPP

International Nuclear Information System (INIS)

Foden, R.W.

1995-01-01

In July 1993 the Commission of the European Communities (CEC) placed a contract with NNC Ltd (National Nuclear Corporation) for performing a Probabilistic Safety Assessment (PSA) for a 1000 MW NPP in the Russian Federation. The contract is part (Project 3.1) of the 1991 TACIS (Technical Assistance to the CIS) programme. This paper describes the objectives and scope of the Project and provides a description of the progress that has been made. For this Project, NNC is the leader of a Consortium of Western European companies that has been formed to undertake this Project and other Projects in the TACIS 91 programme. NNC therefore has overall responsibility for the coordination and management of the complete PSA Project. Other members of the Consortium involved in this Project are Empresarios Agrupados from Spain, Belgatom from Belgium and AEA-Technology from the UK. The analytical work for the Project is performed by the Russian Company Atomenergoproekt in Moscow, under contract to NNC. The official recipient institution for the results of the Project is the Russian Utility, Rosenergatom. The NPP chosen to be the subject of the Project is the Balakovo Unit 4 VVER 1000. (author)

Experience of CR and RCCA operation in Ukrainian WWER-1000: Aspects of reliability, safety and economic efficiency

International Nuclear Information System (INIS)

Afanasyev, A.

2000-01-01

The next topics are represented in the paper: A brief history of WWER-1000 control rod (CR) and WWER-1000 rod cluster control assembly (RCCA) design; Evolution of WWER-1000 CR manufacturing technology and design; Experience of RCCA operation; Lifetime extension of WWER-1000 boron carbide CR; WWER-1000 reactor core operation problems due to partial RCCA insertion; Designing and licensing procedures and first operational experience of WWER-1000 RCCA (CR) with a combined absorber 'boron carbide-hafnium' and a chromium-nickel alloy cladding. The main conclusions are: Fuel assembly (FA) bow is the main reason of partial RCCA insertion during reactor core operation. However, the use of the RCCA and its driver bar with increased dead load, alongside with other measures, allow to reduce the probability of incomplete RCCA insertion; The materials used in CRs of RCCA in existing reactor operating modes have been working reliably; The use of hafnium under an appropriate price policy can give certain economic advantages for the Ukrainian NPPs, however, additional research is needed in order to confirm the specific CR physical characteristics and reliability. (author)

Primary LOCA in VVER-1000 by pressurizer PORV failure

International Nuclear Information System (INIS)

Sabotinov, L.; Lutsanych, S.; Kadenko, I.

2013-01-01

The paper presents the calculations and analysis of the design basis accident of a standard VVER-1000/V320 reactor with inadvertent opening and stuck in open position of the pressurizer pilot operated relief valve (PORV). The objective is the independent assessment of this accident applying the French best estimate thermal-hydraulic computer code CATHARE 2 and verification to meet the safety criteria for such kind of the accident. The 'Inadvertent opening and stuck in open position of PORV' is a design basis accident classified as Medium Break Loss of Coolant Accident (MB LOCA) with the equivalent diameter of the break D- 68 mm. This accident is particularly interesting to be calculated and analyzed, because it took place at operating NPP with VVER-1000 reactors (Rovno NPP) in 2009. The calculations have been carried out with conservative conditions as usual for DBA analysis. The NPP model corresponds to a real VVER-1000/V320 configuration and comprises all safety systems, adopted for one of the latest CATHARE 2 versions. The results of CATHARE 2 calculations are compared with available results of RELAP5 calculations. There is similarity of the thermal-hydraulic parameters behavior, but also some differences can be observed basically due to the break flow prediction, which causes differences in primary pressure evaluation. Both calculations show that there is no boiling crisis in the reactor core and reliable cooldown is achieved. The calculations performed with CATHARE2 code demonstrate the ability of the code to predict reasonably the break flow, pressures, temperatures etc. for considered LOCA scenario and to be applied for safety studies

On integration and innovation of sino-foreign safety culture in Haiyang AP1000 Project

International Nuclear Information System (INIS)

Li Ruipu; Song Fengwei

2010-01-01

The undergoing Haiyang Nuclear Power Plant is not only introducing the top-advanced AP1000 nuclear technology, but also the mature HSE management system from U.S.A. It's very important for both sides to communicate, comprehend and acculturation of both different culture. After over 1 year discussion and practice, the experts of Westinghouse Consortium and Chinese HSE engineers have established an distinctive safety culture of AP1000 Project initially, demonstrating the followings: Exemplary actions of the expat experts and the SNPTC leaders, the high level standard HSE procedures, HSE audit, various training, HSE inspection all-around, the safety performance assessment for prospective index, JHA/JSA , emergency system, humanism rewards and punishment etc.. Haiyang SPMO has made Three-Step master plan for AP1000 project HSE Routine by analysis the site problems and the difference between Chinese and American, that is, from 2008 to 2020, when nuclear power achieve to independent, safety culture of Haiyang AP1000 will change from 'dependent' to 'independent', until the last 'interdependent'. (authors)

Validation cases of CATHARE 2 for VVER-1000 main steam line break analysis

International Nuclear Information System (INIS)

Kolev, Nikolay P.; Petrov, Nikolay; Donov, Jordan; Sabotinov, Luben; Nikonov, Sergey

2008-01-01

Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plan transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation. (author)

Last improvements of the DTC 1000 spectrophotometer

International Nuclear Information System (INIS)

Blanc, F.

1988-01-01

The DTC 1000 spectrophotometer, based on optical fibers and photodiode array was developed for Pu(IV) measurement in fuel reprocessing. The apparatus takes advantage of experience acquired in optical fibers instrumentation and of the study of gamma radiation effects on fibers. This paper give the complete evolution of these last months (memory, hardware, software, background noise). But the principle of the apparatus and examples of results obtained on americium and plutonium on a previous version of the apparatus are recalled. 17 figs., 6 refs [fr

All-cause mortality in asymptomatic persons with extensive Agatston scores above 1000.

Science.gov (United States)

Patel, Jaideep; Blaha, Michael J; McEvoy, John W; Qadir, Sadia; Tota-Maharaj, Rajesh; Shaw, Leslee J; Rumberger, John A; Callister, Tracy Q; Berman, Daniel S; Min, James K; Raggi, Paolo; Agatston, Arthur A; Blumenthal, Roger S; Budoff, Matthew J; Nasir, Khurram

2014-01-01

Risk assessment in the extensive calcified plaque phenotype has been limited by small sample size. We studied all-cause mortality rates among asymptomatic patients with markedly elevated Agatston scores > 1000. We studied a clinical cohort of 44,052 asymptomatic patients referred for coronary calcium scans. Mean follow-up was 5.6 years (range, 1-13 years). All-cause mortality rates were calculated after stratifying by Agatston score (0, 1-1000, 1001-1500, 1500-2000, and >2000). A multivariable Cox regression model adjusting for self-reported traditional risk factors was created to assess the relative mortality hazard of Agatston scores 1001 to 1500, 1501 to 2000, and >2000. With the use of post-estimation modeling, we assessed for the presence of an upper threshold of risk with high Agatston scores. A total of 1593 patients (4% of total population) had Agatston score > 1000. There was a continuous graded decrease in estimated 10-year survival across increasing Agatston score, continuing when Agatston score > 1000 (Agatston score 1001-1500, 78%; Agatston score 1501-2000, 74%; Agatston score > 2000, 51%). After multivariable adjustment, Agatston scores 1001 to 1500, 1501 to 2000, and >2000 were associated with an 8.05-, 7.45-, and 13.26-fold greater mortality risk, respectively, than for Agatston score of 0. Compared with Agatston score 1001 to 1500, Agatston score 1501 to 2000 had a similar all-cause mortality risk, whereas Agatston score > 2000 had an increased relative risk (Agatston score 1501-2000: hazard ratio [HR], 1.01 [95% CI, 0.67-1.51]; Agatston score > 2000: HR, 1.79 [95% CI, 1.30-2.46]). Graphical assessment of the predicted survival model suggests no upper threshold for risk associated with calcified plaque in coronary arteries. Increasing calcified plaque in coronary arteries continues to predict a graded decrease in survival among patients with extensive Agatston score > 1000 with no apparent upper threshold. Published by Elsevier Inc.

25 CFR 1000.464 - What personal conflicts of interest must the standards of conduct regulate?

Science.gov (United States)

2010-04-01

... conduct regulate? 1000.464 Section 1000.464 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS... interest must the standards of conduct regulate? The personal conflicts of interest standards must: (a... financial interest or an employment relationship; (b) Prohibit such officers, employees, or agents from...

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

Energy Technology Data Exchange (ETDEWEB)

Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

2005-07-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Interface between Core/TH Model and Simulator for OPR1000

Energy Technology Data Exchange (ETDEWEB)

Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk; Lee, Seung Ho; Suh, Jung Kwan [KEPRI, Daejeon (Korea, Republic of)

2009-05-15

OPR1000 simulator for ShinKori-Unit 1, which will be operated at 2815MWt of thermal core power, is being developed while the ShinKori-Unit 1 and 2 is being built. OPR1000 simulator adopted the RELAP5 R/T code, which is the adaptation of RELAP5 and NESTLE codes to run in real-time mode with graphical visualization, to model Nuclear Steam Supply System (NSSS) Thermal-Hydraulics (TH) and Reactor Core. The RELAP5 is an advanced, best estimate, reactor TH simulation code developed at Idaho National Engineering and Environment Laboratory(INEEL) and the NESTLE is a true two-energy group neutronics code that computes the neutron flux and power for each node at every time step. As a simulator environment, the 3KEYMASTER{sup TM}, a commercial environment tool of WSC is used.

Interface between Core/TH Model and Simulator for OPR1000

International Nuclear Information System (INIS)

Hwang, Do Hyun; Lee, Myeong Soo; Hong, Jin Hyuk; Lee, Seung Ho; Suh, Jung Kwan

2009-01-01

OPR1000 simulator for ShinKori-Unit 1, which will be operated at 2815MWt of thermal core power, is being developed while the ShinKori-Unit 1 and 2 is being built. OPR1000 simulator adopted the RELAP5 R/T code, which is the adaptation of RELAP5 and NESTLE codes to run in real-time mode with graphical visualization, to model Nuclear Steam Supply System (NSSS) Thermal-Hydraulics (TH) and Reactor Core. The RELAP5 is an advanced, best estimate, reactor TH simulation code developed at Idaho National Engineering and Environment Laboratory(INEEL) and the NESTLE is a true two-energy group neutronics code that computes the neutron flux and power for each node at every time step. As a simulator environment, the 3KEYMASTER TM , a commercial environment tool of WSC is used

Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

International Nuclear Information System (INIS)

Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

2005-01-01

Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

16 CFR 1000.8 - Meetings and hearings; public notice.

Science.gov (United States)

2010-01-01

... FUNCTIONS § 1000.8 Meetings and hearings; public notice. (a) The Commission may meet and exercise all its... agency, including the rendering of official decisions, are generally announced in advance and open to the... appropriate to its functions anywhere in the United States. It will publish a notice of any proposed hearing...

Significant advantages of the safety-first concept in construction, operation, and maintenance of the Westinghosue AP1000 reactor

International Nuclear Information System (INIS)

Cummins, E.; Benitz, K.

2004-01-01

In June 2003, the U.S. Nuclear Regulatory Commission (USNRC) published a draft opinion about safety of the AP1000 Westinghouse pressurized water reactor with 'passive safety' features. The report constitutes an important milestone in the development of the next generation of safe and cost-efficient nuclear power plants. A new AP1000 can be absolutely competitive with fossil fired power plants and may be able to revive the construction of new nuclear power plants worldwide. The reason for designing the AP1000 were safety considerations. The use of passive safety systems at the same time entails a considerable reduction in the costs of design, maintenance, and operation of an AP1000 plant. Independent experts confirmed that an AP1000 can be erected within three years or even less. The estimated electricity generating costs of an AP1000 plant in the United States amount to US Cent 3.2 to 3.6 per kilowatthour. (orig.)

Accidents during shutdown conditions for NPP with WWER-1000/428

International Nuclear Information System (INIS)

Antropov, H.A.

1996-01-01

This paper considers the effect of such internal initiating events upon safety of the NPP with WWER-1000/428 as boron dilution and loss of residual heat removal at the reduced coolant inventory in the primary circuit. 2 refs

Modernization of WWER-1000 radiation monitoring systems

Energy Technology Data Exchange (ETDEWEB)

Smith, T [General Atomics, San Diego, CA (United States)

1996-12-31

A modernization scheme of the radiation monitoring system for WWER-1000 is proposed. It has a purpose to comply with international standards and to reduce operational and maintenance cost by deleting obsolete components and reducing the number of detector channels. Detailed layouts of I/C system architecture, digital radiation monitoring system (DRAMS) architecture and LRP block diagram are presented. If planned and implemented properly, this program can provide cost savings by reducing time required to access and display data and maintenance cost by deleting obsolete parts and decreasing the number of detector channels. 3 figs.

Modernization of WWER-1000 radiation monitoring systems

International Nuclear Information System (INIS)

Smith, T.

1995-01-01

A modernization scheme of the radiation monitoring system for WWER-1000 is proposed. It has a purpose to comply with international standards and to reduce operational and maintenance cost by deleting obsolete components and reducing the number of detector channels. Detailed layouts of I/C system architecture, digital radiation monitoring system (DRAMS) architecture and LRP block diagram are presented. If planned and implemented properly, this program can provide cost savings by reducing time required to access and display data and maintenance cost by deleting obsolete parts and decreasing the number of detector channels. 3 figs

25 CFR 1000.382 - What may the Tribe's/Consortium's annual report on self-governance address?

Science.gov (United States)

2010-04-01

...-governance address? 1000.382 Section 1000.382 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS... report on self-governance address? (a) The Tribe's/Consortium's annual self-governance report may address... the programs and services funded under self-governance, summarized and annotated as the Tribe may deem...

Investigation of mutations in the HBB gene using the 1,000 genomes database.

Science.gov (United States)

Carlice-Dos-Reis, Tânia; Viana, Jaime; Moreira, Fabiano Cordeiro; Cardoso, Greice de Lemos; Guerreiro, João; Santos, Sidney; Ribeiro-Dos-Santos, Ândrea

2017-01-01

Mutations in the HBB gene are responsible for several serious hemoglobinopathies, such as sickle cell anemia and β-thalassemia. Sickle cell anemia is one of the most common monogenic diseases worldwide. Due to its prevalence, diverse strategies have been developed for a better understanding of its molecular mechanisms. In silico analysis has been increasingly used to investigate the genotype-phenotype relationship of many diseases, and the sequences of healthy individuals deposited in the 1,000 Genomes database appear to be an excellent tool for such analysis. The objective of this study is to analyze the variations in the HBB gene in the 1,000 Genomes database, to describe the mutation frequencies in the different population groups, and to investigate the pattern of pathogenicity. The computational tool SNPEFF was used to align the data from 2,504 samples of the 1,000 Genomes database with the HG19 genome reference. The pathogenicity of each amino acid change was investigated using the databases CLINVAR, dbSNP and HbVar and five different predictors. Twenty different mutations were found in 209 healthy individuals. The African group had the highest number of individuals with mutations, and the European group had the lowest number. Thus, it is concluded that approximately 8.3% of phenotypically healthy individuals from the 1,000 Genomes database have some mutation in the HBB gene. The frequency of mutated genes was estimated at 0.042, so that the expected frequency of being homozygous or compound heterozygous for these variants in the next generation is approximately 0.002. In total, 193 subjects had a non-synonymous mutation, which 186 (7.4%) have a deleterious mutation. Considering that the 1,000 Genomes database is representative of the world's population, it can be estimated that fourteen out of every 10,000 individuals in the world will have a hemoglobinopathy in the next generation.

Comparative analysis of exercise 2 results of the OECD WWER-1000 MSLB benchmark

International Nuclear Information System (INIS)

Kolev, N.; Petrov, N.; Royer, E.; Ivanov, B.; Ivanov, K.

2006-01-01

In the framework of joint effort between OECD/NEA, US DOE and CEA France a coupled three-dimensional (3D) thermal-hydraulic/neutron kinetics benchmark for WWER-1000 was defined. Phase 2 of this benchmark is labeled W1000CT-2 and consists of calculation of a vessel mixing experiment and main steam line break (MSLB) transients. The reference plant is Kozloduy-6 in Bulgaria. Plant data are available for code validation consisting of one experiment of pump start-up (W1000CT-1) and one experiment of steam generator isolation (W1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of the available results for W1000CT-2 Exercise 2 devoted to core-vessel calculation with imposed MSLB vessel boundary conditions. Because of the recent re-calculation of the cross-section libraries, core physics results from PARCS and CRONOS codes could be compared only. The comparison is code-to-code (including BIPR7A/TVS-M lib) and code vs. plant measured data in a steady state close to the MSLB initial state. The results provide a test of the cross-section libraries and show a good agreement of plant measured and computed data. The comparison of full vessel calculations was made from the point of view of vessel mixing, considering mainly the coarse-mesh features of the flow. The FZR and INRNE results from multi-1D calculations with different mixing models are similar, while the FZK calculations with a coarse-3D vessel model show deviations from the others. These deviations seem to be due to an error in the use of a boundary condition after flow reversal (Authors)

24 CFR 1000.120 - May a recipient use Indian preference or tribal preference in selecting families for housing...

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false May a recipient use Indian preference or tribal preference in selecting families for housing assistance? 1000.120 Section 1000.120... AMERICAN HOUSING ACTIVITIES Affordable Housing Activities § 1000.120 May a recipient use Indian preference...

24 CFR 1000.162 - How will a recipient know that non-dwelling structures assisted under the IHBG program meet the...

Science.gov (United States)

2010-04-01

...-dwelling structures assisted under the IHBG program meet the requirements of 1000.160? 1000.162 Section... that non-dwelling structures assisted under the IHBG program meet the requirements of 1000.160? (a) The... non-dwelling structure meets the limitation requirements of § 1000.160. If the recipient develops...

The application of ZED-2 data in the design and analysis of the ACR-1000

International Nuclear Information System (INIS)

Dweiri, Y.

2011-01-01

The presentation discusses how data obtained from experiments performed in the ZED-2 Critical Facility at Atomic Energy of Canada Limited's Chalk River Laboratories is used in the design and analysis of the ACR-1000. The ACR-1000 design is an evolutionary advancement of the well-established CANDU Reactor technology to Generation 3+. When key core physics parameters for the ACR-1000 are calculated, biases and uncertainties in using the reactor physics toolset WIMS AECL/DRAGON/RFSP and MCNP must be quantified to be able to appropriately assess safety margins in the design. A unique feature of ZED-2 is the ability to perform a wide variety of experiments which is important in providing data that is used in the design and analysis of the ACR-1000. The ACR-1000 was designed to have a small and slightly negative coolant-void-reactivity (CVR), which provides an inherent safety characteristic to mitigate the reactivity transient under any postulated loss-of-coolant accident (LOCA) scenarios. The bias and uncertainty for CVR (full and checkerboard) calculations, derived through comparisons with experimental data, is pivotal to assess the accuracy of the predicted CVR value. This bias and uncertainty is also used in important LOCA analysis to demonstrate the safety features of the ACR-1000. This crucial physics information is available through experiments performed in ZED-2. Other important data that is obtained from ZED-2 in assessing core design pertains to end-flux peaking and in-core flux detector response. The end-region configuration of ACR fuel bundles is an important component of the design, as the axial power distribution within the bundle is not uniform. Detector response in an ACR flux spectrum is important to determine the suitability of safety system detectors when used in the ACR-1000 core. These applications demonstrate the versatility of ZED-2 in providing important reactor physics data makes this critical facility indispensible in current and future R and D

Prezident rekomendujet podnjat detskoje posobije do 1000 kron v mesjats

Index Scriptorium Estoniae

2002-01-01

Emadepäevatervituses toetas president Aino Järvesoo algatust tõsta lastetoetusi 1000 kroonini kuus. Ilmunud ka Narvskaja Gazeta (2002/May/14) lk. 1. Pealk. ajalehes Sillamjaeskii Vestnik (2002/May/14): "Prezident zajavil, tshto javljajetsja storonnikom uvelitshenija detskogo posobija", lk. 1

Evaluation of WWER-1000 vessel materials fracture toughness

International Nuclear Information System (INIS)

Grinik, Eh.U.; Revka, V.N.; Chirko, L.I.; Chajkovskij, Yu.V.

2007-01-01

The lifetime of WWER-1000-type reactor vessels is finally conditioned by the fracture toughness (crack growth resistance) of RPV materials. Up to now in line with the regulations the fracture toughness is characterized by the critical temperature of brittleness determined by the results of the Charpy specimen impact testing. Such approach is typical for all countries operating the water pressure reactors. However, regulatory approach is known from the western specialists not always to characterize adequately the crack growth resistance of the vessel materials and in some cases to underestimate their characteristics in the reference state that leads to unreasonably high conservatism. Excessive conservatism may lead to the invalid restrictions in the operating modes and the service life of the reactor vessel. Therefore there appeared the necessity to apply another approaches based on the state-of-the-art experimental methods of the fracture mechanics and allowing evaluating the fracture toughness parameters sufficiently. The paper presents the results of the comparison of the regulatory approach and the Master curve approach from the point of view of the adequate determination of the vessel material crack growth resistance parameters. Analysis of the experimental data of the surveillance specimens illustrated the potential possibility of applying the new statistical method for the WWER-1000- type reactor vessel lifetime extension

1000 MW steam turbine for nuclear power station

International Nuclear Information System (INIS)

Drahy, J.

1987-01-01

Skoda Works started the manufacture of the 1000 MW steam turbine for the Temelin nuclear power plant. The turbine will use saturated steam at 3,000 r.p.m. It will allow steam supply to heat water for district heating, this of an output of 893 MW for a three-stage water heating at a temperature of 150/60 degC or of 570 MW for a two-stage heating at a temperature of 120/60 degC. The turbine features one high-pressure and three identical low-pressure stages. The pressure gradient between the high-pressure and the low-pressure parts was optimized as concerns the thermal efficiency of the cycle and the thermodynamic efficiency of the low-pressure part. A value of 0.79 MPa was selected corresponding to the maximum flow rate of the steam entering the turbine. This is 5,495 t/h, the admission steam parameters are 273.3 degC and 5.8 MPa. The feed water temperature is 220.9 degC. It is expected that throughout the life of the turbine, there will be 300 cold starts, 1,000 starts following shutdown for 55 to 88 hours, and 600 starts following shutdown for 8 hours. (Z.M.). 8 figs., 1 ref

Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

International Nuclear Information System (INIS)

Dye, M.; Shah, H.

2015-01-01

Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

Main trends of upgrading the 1000 MW steam turbine

International Nuclear Information System (INIS)

Drahy, J.

1990-01-01

Parameters are compared for the 1000 MW steam turbine manufactured by the Skoda Works, Czechoslovakia, and turbines in the same power range by other manufacturers, viz. ABB, Siemens/KWU, GEC and LMZ. The Skoda turbine compares well with the other turbines with respect to all design parameters, and moreover, enables the most extensive heat extraction for district heating purposes. The main trends in upgrading this turbine are outlined; in particular, they include an additional increase in the heat extraction, which is made possible by a new design of the low-pressure section or by using a ''satellite'' turbine. The studies performed also indicate that the output of the full-speed saturated steam turbine can be increased to 1300 MW. An experimental turbine representing one flow of the high-pressure part of the 1000 MW turbine is being built on the 1:1 scale. It will serve to verify the methods of calculation of the wet steam flow and to experimentally test the high-pressure part over a wide span of the parameters. (Z.M.). 1 tab., 3 figs., 7 refs

OPR1000 RCP Flow Coastdown Analysis using SPACE Code

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

Development of UO{sub 2}-Stainless Steel Fuel Plates Containing 30-50 Vol. % Oxide; Fabrication de plaques de combustible en acier inoxydable-UO{sub 2} contenant 30 a 40% d'oxyde (en volume); Razrabotka toplivnykh ehlementov iz nerzhaveyushchej stali i UO{sub 2}, soderzhashchikh 30 - 50 OB.% okisi; Elaboracion de placas de combustible de acero inoxidable UO{sub 2} conteniendo 30 a 40% de oxido (en volumen)

Energy Technology Data Exchange (ETDEWEB)

Lloyd, H. [Atomic Energy Research Establishment, Harwell (United Kingdom)

1963-11-15

de nucleos de cermets constituidos por placas con las particulas uniformemente distribuidas en la matriz de acero inoxidable, asi como procedimientos para unir las placas de combustible por laminacion en caliente. Se ha estudiado la laminacion a temperaturas de hasta 1300{sup o}C, utilizando deformaciones totales del orden del 40 al 90%, para determinar las condiciones optimas de produccion de nucleos de elevada densidad y obtener una union satisfactocia entre los componentes de las placas, con fragmentacion y estiraje minimos de las particulas de UO{sub 2} . Tambien se describe la fabricacion de grandes placas de combustible utilizando materiales polinucleares que se unen por laminacion en caliente. Se facilitan datos sobre las propiedades mecanicas de cermets de UO{sub 2} y acero inoxidable de 30, 40 y 50% en volumen, elaborados como se ha indicado y ensayados tal como quedan despues de laminarlos, y despues de someterlos a recocido a diversas temperaturas que alcanzaron hasta 700{sup o}C, utilizando probetas tomadas transversal y longitudinalmente con respecto a la direccion de laminado. Se examina la influencia del tamaflo y uniformidad de la distribucion de las esferas de UO{sub 2} en la estabilidad de las propiedades mecanicas. Tambien se evalua la solidez de la union entre el nucleo y el revestimiento de cermets del mismo tipo, en identico intervalo de temperaturas. Asimismo, se incluyen los resultados de los ensayos de ciclado termico efectuados entre 50 y 800{sup o}C con objeto de estudiar los efectos en la estructura del cermet y en la estabilidad de la union, despues de 100, 500 y 1000 ciclos. (author) [Russian] Daetsya opisanie protsessov izgotovleniya toplivnykh ehlementov plastinchatogo tipa iz UO{sub 2} i nerzhaveyushchej stali, soderzhashchikh do 50 ob% UO{sub 2}. Opisyvaetsya izgotovlenie ochen' plotnykh spekshikhsya sharoobraznykh chastits iz UO{sub 2} v intervale razmerov ot 100 do 500, a takzhe pressovanie i spekanie kermetnykh plastinchatykh

Nizek površinski stanovanjski standard v Sloveniji: nizka kupna moč kot večno opravičilo

Directory of Open Access Journals (Sweden)

Richard Sendi

2013-01-01

Full Text Available V prispevku predstavljamo in zagovarjamo tezo, da ima Slovenija nizek stanovanjski standard in da večina prebivalcev živi v stanovanjih, ki jih je po mednarodno uveljavljenih standardih mogoče opredeliti za prenaseljena. Resničnost te teze dokazujemo s statističnimi podatki ter tudi s pomočjo pregleda literature in predstavitvijo stališč različnih domačih avtorjev, ki so obravnavali to temo. Poleg obširnega pregleda tuje in domače literature na to temo predstavljamo tudi pregled zgodovinskih dejavnikov, za katere ocenjujemo, da so v preteklosti pomembno vplivali na vzpostavitev in uveljavitev današnjih stanovanjskih standardov. V sklepnem delu posebej poudarjamo danes prevladujoče napačno stališče, ki ga opredeljujemo kot najpomembnejši zaviralni dejavnik pri razmišljanjih o potrebnem izboljšanju slovenskih stanovanjskih standardov. V tem delu podajamo nove teoretične razlage o tem, zakaj se v Sloveniji še vedno vzdržuje nizek površinski stanovanjski standard. Končni namen prispevka pa je sprožiti izvedbo dejanskih ukrepov, ki bodo pripeljali do uveljavljanja ustreznejših standardov in izboljšanja stanja.

[Experimental evaluation of the Sysmex UF-1000i for ruling out non-gonococcal urethritis].

Science.gov (United States)

Grosso, Shamanta; Bruschetta, Graziano; Camporese, Alessandro

2012-09-01

Acute nongonococcal urethritis (NGU) is one of the commonest sexually transmitted infections affecting men and women. The diagnosis of NGU has traditionally required microscopic evidence of urethritis. However, a significant proportion of patients with urethral symptoms do not have microscopic evidence of urethritis. The purpose of the present study was to evaluate the analytical performance of the UF1000i, a recently introduced fluorescence flow cytometer intended for urinalysis purposes which provides new analytical features that seem particularly suitable for microbiological diagnostics, for ruling out NGU or predicting the presence of infection. The Sysmex UF1000i is a flow cytometry analyzer capable of quantifying a lot of particles, including bacteria (BACT) and white blood cells (WBCs). To evaluate the analytical performance of the UF1000i as a method for ruling out NGU, we examined 200 urethral smear samples, collected in a new liquid transport medium (Copan), and compared the UF1000i results with standard culture/molecular and microscopic Gram stain results. With instrument cut-off values of 200 BACT x 10^6/L and 500 WBCs x 10^6/L, we obtained a sensitivity of 84%, a specificity of 82%, and a high negative predictive value (96%). Culture/molecular detection of pathogens remains the gold standard technique for the diagnosis of NGU. However, the Sysmex UF1000i is capable of improving the efficiency of NGU presumptive diagnosis, providing results in a few minutes, with a high negative predictive value and high values of sensitivity.

Westinghouse and nuclear renaissance. The Westinghouse AP1000 - a technology solution for Slovakia

International Nuclear Information System (INIS)

Kirst, M.

2009-01-01

The Westinghouse AP1000 nuclear reactor design has been chosen by both China and the United States as the preferred technology in their new reactor programs. With four reactors in China and six in the United States under contract, in addition to the only Generation III+ design with NRC certification as well as the European Utility Requirements certification, the AP1000 has both a strong global customer base and regulatory certainty to facilitate its adoption in the Slovak Republic. (author)

21 CFR 1000.50 - Recommendation for the use of specific area gonad shielding on patients during medical diagnostic...

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Recommendation for the use of specific area gonad shielding on patients during medical diagnostic x-ray procedures. 1000.50 Section 1000.50 Food and Drugs... GENERAL Radiation Protection Recommendations § 1000.50 Recommendation for the use of specific area gonad...

24 CFR 1000.238 - What percentage of the IHBG funds can be used for administrative and planning expenses?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What percentage of the IHBG funds can be used for administrative and planning expenses? 1000.238 Section 1000.238 Housing and Urban... ACTIVITIES Indian Housing Plan (IHP) § 1000.238 What percentage of the IHBG funds can be used for...

PENGARUH KONSENTRASI BAHAN STABILISATOR PEG 1000 DAN LAMA PERENDAMAN TERHADAP KESTABILAN DIMENSI KAYU KECAPI (Sandoricum koetjape Merr

Directory of Open Access Journals (Sweden)

Lusyiani Lusyiani

2016-11-01

Full Text Available Kayu bersifat higroskopis dan anisotropis, karena itulah untuk meningkatkan kualitas kayu diperlukan suatu perlakuan stabilisasi dimensi.  Tujuan dari penelitian ini adalah untuk menguji pengaruh konsentrasi Polyethylene Glycol 1000 dan waktu perendaman terhadap stabilisasi dimensi kayu yakni nilai penyusutannya.  Sampel yang digunakan dalam penelitian ini adalah Polyethylene glycol 1000, kayu Kecapi dan air suling sebagai pelarutnya.  Proses stabilisasi dimensi dilakukan dengan menggunakan metode proses difusi bulking yang mengganti bahan penstabilnya dengan air.  Sampel yang digunakan dalam penelitian ini adalah sebanyak 120 buah sampel.  Hasil penelitian menunjukkan bahwa Polyethylene Glycol 1000 pada tingkat konsentrasi 30% dan lama perendaman 5 hari memiliki nilai stabilisasi dimensi terbaik untuk semua parameter yang diuji.  Tingkat konsentrasi tertinggi untuk Polyethylene Glycol 1000 dan lama perendaman yang terlama, nilai penyusutannya paling kecil dan juga dapat meningkatkan nilai stabilisasi dimensi. Wood is hygroscopic and anisotropic, hence it is needed to improve the quality of wood by dimension stabilization treatment. The aim of study is to examine the effect of Polyethylene Glycol  1000 concentration and soaking time on the shrinkage rate stabilization dimensions. Polyethylene glycol 1000 and Kecapi wood were used in the research and distilled water as a solvent.  Stabilization was carried out by bulking method that replaced stabilizer material with water by diffusion process.  A member of 120 pieces samples were used in the study. The research results showed that Polyethylene Glycol  1000 at 30% concentration level and 5 days of soaking time have the best value of stabilization for all tested parameters.  The hingher level concentration of Polyethylene Glycol 1000 and the longer immersion time, smaller shrinkage and increasing the value of stabilization dimensions, as well.

The acquisition system for Tore Supra 1000 s discharges

International Nuclear Information System (INIS)

Guillerminet, B.; How, J.

2000-01-01

Long duration discharges are planned for Tore Supra in the near future. A study has been made to detect and correct all the possible limitations of the data acquisition system. Results and analysis of a few 1000 s 'dry-run' test pulses are presented in this paper as well as the solutions foreseen for Tore Supra

Determination of fast neutron fluence at WWER-1000 pressure vessel

International Nuclear Information System (INIS)

Valenta, V. et al.

1989-01-01

The influence function method is an effective tool making it possible, by means of tabulated values to rapidly perform three-dimensional calculations of fast neutron fluences for various reactor core loadings and for various nuclear power plant units. The procedure for determining the spatial dependence of the fast neutron fluences in a WWER-1000 pressure vessel is described. For this, the reactor core is divided into sufficiently fine volume elements within which the neutron source can be regarded as coordinate-independent. The influence functions point to a substantial role of sources lying at the reactor core periphery. In WWER-1000 reactors, only 1 or 2 rows of peripheral assemblies are important. The influence function method makes possible a rapid and easy determination of preconditions for the assessment of the residual lifetime of the pressure vessel based on the actual reactor core loadings. (Z.M.). 7 figs., 8 refs

AP1000R design robustness against extreme external events - Seismic, flooding, and aircraft crash

International Nuclear Information System (INIS)

Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J.

2012-01-01

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000 R nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is

Impact of cross-section generation procedures on the simulation of the VVER 1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3-D thermal hydraulics/ neutron kinetics models

International Nuclear Information System (INIS)

Boyan D Ivanov; Kostadin N Ivanov; Sylvie Aniel; Eric Royer

2005-01-01

Full text of publication follows: In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D thermal hydraulics/neutron kinetics benchmark was defined. The overall objective OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3-D thermal hydraulics/ neutron kinetics models based on the data available in the benchmark specifications. The first code to code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between 2 sets of results, one of them being close to experimental results. The present paper focuses first on the analysis of the observed discrepancies. The VVER 1000 3-D thermal hydraulics/neutron kinetics models are based on thermal-hydraulic and neutronic data homogenized at the assembly scale. The neutronic data, provided as part of the benchmark specifications, consist thus in a set of parametrized 2 group cross sections libraries representing the different assemblies and the reflectors. The origin of the high observed discrepancies was found to lie in the use of these neutronic libraries. The concern was then to find a way to provide neutronic data, compatible with all the benchmark participants neutronic models, that enable also comparisons with experimental results. An analysis of the

Dosimetry of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens as a part of PLiM at Ukrainian NPPs

International Nuclear Information System (INIS)

Bukanov, V.N.; Diemokhin, V.L.; Grytsenko, O.V.; Ilkovych, V.V.; Pugach, A.M.; Pugach, S.M.; Vasylieva, O.G.; Vyshnevskyi, I.M.; Kasatkin, O.G.

2012-01-01

A regular surveillance program for VVER-1000 and its shortages are described. The Methodology for determination of neutron flux functionals on surveillance specimens of VVER-1000 pressure vessel is presented. The radiation exposure monitoring system for VVER-1000 pressure vessel is described. The main principles of an additional surveillance program for VVER-1000 are presented. The Dosimetry Experiment, which is already carrying out at Unit 3 of Rivne NPP, is described. (author)

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

16 CFR 1000.23 - Office of Information and Technology Services.

Science.gov (United States)

2010-01-01

... COMMISSION ORGANIZATION AND FUNCTIONS § 1000.23 Office of Information and Technology Services. The Office of Information and Technology Services houses the Commission's Secretariat, which facilitates the preparation of... agency's Chief Freedom of Information Act Officer. The Office of Information and Technology Services is...

29 CFR 780.1000 - Scope and significance of interpretative bulletin.

Science.gov (United States)

2010-07-01

... STANDARDS ACT Employment of Home- workers in Making Wreaths; Exemption From Minimum Wage, Overtime... minimum wage, overtime pay, and child labor provisions of the Act for certain homeworkers employed in... Section 780.1000 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF...

Uncertainty Evaluation with Multi-Dimensional Model of LBLOCA in OPR1000 Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Jieun; Oh, Deog Yeon; Seul, Kwang-Won; Lee, Jin Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

KINS has used KINS-REM (KINS-Realistic Evaluation Methodology) which developed for Best- Estimate (BE) calculation and uncertainty quantification for regulatory audit. This methodology has been improved continuously by numerous studies, such as uncertainty parameters and uncertainty ranges. In this study, to evaluate the applicability of improved KINS-REM for OPR1000 plant, uncertainty evaluation with multi-dimensional model for confirming multi-dimensional phenomena was conducted with MARS-KS code. In this study, the uncertainty evaluation with multi- dimensional model of OPR1000 plant was conducted for confirming the applicability of improved KINS- REM The reactor vessel modeled using MULTID component of MARS-KS code, and total 29 uncertainty parameters were considered by 124 sampled calculations. Through 124 calculations using Mosaique program with MARS-KS code, peak cladding temperature was calculated and final PCT was determined by the 3rd order Wilks' formula. The uncertainty parameters which has strong influence were investigated by Pearson coefficient analysis. They were mostly related with plant operation and fuel material properties. Evaluation results through the 124 calculations and sensitivity analysis show that improved KINS-REM could be reasonably applicable for uncertainty evaluation with multi-dimensional model calculations of OPR1000 plants.

Performance evaluation of XN-1000 Sysmex blood cell count of low value of white blood cell%Sysmex XN-1000血细胞分析仪计数低值白细胞性能的评估

Institute of Scientific and Technical Information of China (English)

蒋清清; 马萍; 丁爽

2016-01-01

目的：探讨 Sysmex XN‐1000（后简称 XN‐1000）全自动血细胞分析仪检测白细胞计数（WBC）及分类的准确性。方法收集该院2015年1～4月门诊及住院患者及健康体检者共206例，根据人工镜检检测 WBC 结果分为高值组（WBC ＞10．0×109／L），健康组（4．0×109／L ＜ WBC ≤10．0×109／L）和低值组（1．5×109／L ＜ WBC ≤4．0×109／L）。再分别用 XN‐1000与Sysmex XE‐5000（后简称 XE‐5000）进行 WBC 并分类。采用统计学软件对3种方法的结果进行对比分析，并进行相关性分析。结果3种方法检测 WBC 结果比较，高值组和健康组差异无统计学意义（P＞0．05），结果具有正相关性。低值组中，XN‐1000检测结果准确性与 XE‐5000差异无统计学意义（P＞0．05），但是白细胞分类结果显著高于 XE‐5000（P＜0．05）；对于异常细胞，XN‐1000和 XE‐5000不能进行有效分类，仍需人工镜检进行分类。结论 XN‐1000全自动血细胞分析仪在 WBC 与 XE‐5000及人工镜检比较无差异统计学意义，在测定白细胞分类上 XN‐1000较 XE‐5000更准确（XN‐1000合格率91％，XE‐5000合格率63％），但仍不能完全替代人工镜检。对于异常细胞，XN‐1000不能有效分类，需与人工镜检结合。%Objective To investigate the accuracy of XN‐1000 automatic blood cell analyzer in detecting of white blood cell (WBC) count and classification .Methods A total of 206 cases including outpatients ,inpatients and health people were collected in the hospital from January to April 2015 .According to manual microscopy detection of white blood cell count results ,the cases were divided into three groups :high value of white blood cell(WBC was more than 10 .0 × 109 /L) group ,normal value of white blood cells group(WBC was more than 4 .0 × 109 /L but less than 10 .0 × 109 /L) and low value of white blood cell(WBC was more than 1 .5 × 109 /L

25 CFR 1000.173 - How does a newly selected Tribe/Consortium initiate the negotiation phase?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false How does a newly selected Tribe/Consortium initiate the negotiation phase? 1000.173 Section 1000.173 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATIO...

PI-RADS version 2: evaluation of diffusion-weighted imaging interpretation between b = 1000 and b = 1500 s mm-2.

Science.gov (United States)

Kwon, Mi-Ri; Kim, Chan Kyo; Kim, Jae-Hun

2017-11-01

To investigate the variability of diffusion-weighted imaging (DWI) interpretation of Prostate Imaging Reporting and Data System (PI-RADS) version 2 (v2) in evaluating prostate cancer (PCa). 154 patients with PCa underwent multiparametric 3T MRI, followed by radical prostatectomy. DWI with different b values (b = 0, 100, 1000 and 1500 s mm - 2 ) was obtained. Using the PI-RADS v2, two radiologists independently scored suspicious lesions in each patient and compared DWI of b = 1000 (DWI 1000 ) with 1500 (DWI 1500 ) s mm - 2 . On DWI 1000 and DWI 1500 , the intermethod and interobserver agreements of DWI scores were excellent in all patients (κ ≥ 0.873). In each peripheral zone and transition zone DWI scores, both observers showed excellent intermethod agreement between DWI 1000 and DWI 1500 (κ ≥ 0.897), and interobserver agreement for DWI 1000 and DWI 1500 was good to excellent (κ ≥ 0.796). For estimating clinically significant cancer, the area under receiver operating characteristics curves of DWI 1000 and DWI 1500 were 0.710 and 0.724 for observer 1 (p = 0.11), and 0.649 and 0.656 for observer 2 (p = 0.12), respectively. The PI-RADS v2 scoring at 3T shows excellent agreement between DWI 1000 and DWI 1500 in evaluating PCa, with excellent inter-observer agreement. Advance in knowledge: DWI using b = 1000 s mm -2 instead of b = 1500 s mm -2 reduces examination time or image distortion, with improved the signal-to-noise ratio.

Advanced fuel cycles of WWER-1000 reactors

International Nuclear Information System (INIS)

Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, A.

2003-01-01

The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

The challenge of reducing energy consumption of the Top-1000 largest industrial enterprises in China

International Nuclear Information System (INIS)

Price, Lynn; Wang Xuejun; Yun Jiang

2010-01-01

In 2005, the Chinese government announced an ambitious goal of reducing energy consumption per unit of gross domestic product (GDP) by 20% between 2005 and 2010. One of the key initiatives for realizing this goal is the Top-1000 Energy-Consuming Enterprises program. The energy consumption of these 1000 enterprises accounted for 33% of national and 47% of industrial energy usage in 2004. Under the Top-1000 program, 2010 energy consumption targets were determined for each enterprise. The objective of this article is to evaluate the program design and initial results, given limited information and data, to understand the possible implications of its success in terms of energy and carbon dioxide emission reductions and to recommend future program modifications based on international experience with similar target-setting agreement programs. Even though the Top-1000 program was designed and implemented rapidly, it appears that - depending upon the GDP growth rate - it could contribute to somewhere between approximately 10% and 25% of the savings required to support China's efforts to meet a 20% reduction in energy use per unit of GDP by 2010.

Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP

International Nuclear Information System (INIS)

Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.

2011-01-01

The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)

Performance of genotype imputation for low frequency and rare variants from the 1000 genomes.

Science.gov (United States)

Zheng, Hou-Feng; Rong, Jing-Jing; Liu, Ming; Han, Fang; Zhang, Xing-Wei; Richards, J Brent; Wang, Li

2015-01-01

Genotype imputation is now routinely applied in genome-wide association studies (GWAS) and meta-analyses. However, most of the imputations have been run using HapMap samples as reference, imputation of low frequency and rare variants (minor allele frequency (MAF) 1000 Genomes panel) are available to facilitate imputation of these variants. Therefore, in order to estimate the performance of low frequency and rare variants imputation, we imputed 153 individuals, each of whom had 3 different genotype array data including 317k, 610k and 1 million SNPs, to three different reference panels: the 1000 Genomes pilot March 2010 release (1KGpilot), the 1000 Genomes interim August 2010 release (1KGinterim), and the 1000 Genomes phase1 November 2010 and May 2011 release (1KGphase1) by using IMPUTE version 2. The differences between these three releases of the 1000 Genomes data are the sample size, ancestry diversity, number of variants and their frequency spectrum. We found that both reference panel and GWAS chip density affect the imputation of low frequency and rare variants. 1KGphase1 outperformed the other 2 panels, at higher concordance rate, higher proportion of well-imputed variants (info>0.4) and higher mean info score in each MAF bin. Similarly, 1M chip array outperformed 610K and 317K. However for very rare variants (MAF ≤ 0.3%), only 0-1% of the variants were well imputed. We conclude that the imputation of low frequency and rare variants improves with larger reference panels and higher density of genome-wide genotyping arrays. Yet, despite a large reference panel size and dense genotyping density, very rare variants remain difficult to impute.

Risk Factors for Childhood Obesity in the First 1,000 Days: A Systematic Review.

Science.gov (United States)

Woo Baidal, Jennifer A; Locks, Lindsey M; Cheng, Erika R; Blake-Lamb, Tiffany L; Perkins, Meghan E; Taveras, Elsie M

2016-06-01

Mounting evidence suggests that the origins of childhood obesity and related disparities can be found as early as the "first 1,000 days"-the period from conception to age 2 years. The main goal of this study is to systematically review existing evidence for modifiable childhood obesity risk factors present from conception to age 2 years. PubMed, Embase, and Web of Science were searched for studies published between January 1, 1980, and December 12, 2014, of childhood obesity risk factors present during the first 1,000 days. Prospective, original human subject, English-language research with exposure occurrence during the first 1,000 days and with the outcome of childhood overweight or obesity (BMI ≥85th percentile for age and sex) collected between age 6 months and 18 years were analyzed between December 13, 2014, and March 15, 2015. Of 5,952 identified citations, 282 studies met inclusion criteria. Several risk factors during the first 1,000 days were consistently associated with later childhood obesity. These included higher maternal pre-pregnancy BMI, prenatal tobacco exposure, maternal excess gestational weight gain, high infant birth weight, and accelerated infant weight gain. Fewer studies also supported gestational diabetes, child care attendance, low strength of maternal-infant relationship, low SES, curtailed infant sleep, inappropriate bottle use, introduction of solid food intake before age 4 months, and infant antibiotic exposure as risk factors for childhood obesity. Modifiable risk factors in the first 1,000 days can inform future research and policy priorities and intervention efforts to prevent childhood obesity. Copyright © 2016 American Journal of Preventive Medicine. Published by Elsevier Inc. All rights reserved.

The Cementation of Boron to Steels by the Method of Electrolytic Deposition; Cementation Electrolytique d'Aciers par le Bore; Tsementirovanie bora v stali putem ehlektroliticheskogo osazhdeniya; Cementacion de Aceros con Boro por Deposito Electrolitico

Energy Technology Data Exchange (ETDEWEB)

Kawakami, M. [Tokyo Institute of Technology (Japan)

1964-06-15

aceros al carbono, aceros inoxidables, acero de elevado contenido de cromo, etc.; todos ellos dan resultados satisfactorios. (author) [Russian] Privoditsja metod izgotovlenija upravljajushhego sterzhnja s cementirovaniem bora. Jetot process osushhestvljaetsja posredstvom jelektroliza rasplavlennogo tetrabornokislogo natrija pri temperature okolo 900 Degree-Sign S . Obrabatyvaemyj stal'noj sterzhen' sluzhit katodom, v k a chestve anoda ispol'zuetsja grafitovyj jelektrod. V processe jelektroliza bor osazhdaetsja na stal'noj sterzhen' i diffundiruet v sta l'. Pri podhodjashhih uslovijah jelektroliza legko udaetsja poluchit' stal'noj sterzhen', soderzhashhij po 20% bora. Issledovalis' sledujushhie tipy stali: uglerodistaja, nerzhavejushhie, hromistaja i t.d. , vo vseh sluchajah byli polucheny polozhitel'nye rezul'taty. (author)

24 CFR 1000.122 - May NAHASDA grant funds be used as matching funds to obtain and leverage funding, including any...

Science.gov (United States)

2010-04-01

... considered an affordable housing activity? 1000.122 Section 1000.122 Housing and Urban Development... Housing Activities § 1000.122 May NAHASDA grant funds be used as matching funds to obtain and leverage...

25 CFR 1000.47 - What must a Tribe/Consortium do to receive a negotiation grant?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What must a Tribe/Consortium do to receive a negotiation grant? 1000.47 Section 1000.47 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Section...

25 CFR 1000.46 - Which Tribes/Consortia may be selected to receive a negotiation grant?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false Which Tribes/Consortia may be selected to receive a negotiation grant? 1000.46 Section 1000.46 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT...

25 CFR 1000.350 - What is the purpose of this subpart?

Science.gov (United States)

2010-04-01

... EDUCATION ACT Trust Evaluation Review § 1000.350 What is the purpose of this subpart? This subpart describes how the trust responsibility of the United States is legally maintained through a system of trust...

JSD1000: multi-group cross section sets for shielding materials

International Nuclear Information System (INIS)

Yamano, Naoki

1984-03-01

A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)

25 CFR 1000.163 - Can a Tribe/Consortium negotiate other terms and conditions not contained in the model compact?

Science.gov (United States)

2010-04-01

... conditions not contained in the model compact? 1000.163 Section 1000.163 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF... Funding Agreements Negotiating A Self-Governance Compact § 1000.163 Can a Tribe/Consortium negotiate other...

Genomic Prediction from Whole Genome Sequence in Livestock: The 1000 Bull Genomes Project

DEFF Research Database (Denmark)

Hayes, Benjamin J; MacLeod, Iona M; Daetwyler, Hans D

Advantages of using whole genome sequence data to predict genomic estimated breeding values (GEBV) include better persistence of accuracy of GEBV across generations and more accurate GEBV across breeds. The 1000 Bull Genomes Project provides a database of whole genome sequenced key ancestor bulls....... In a dairy data set, predictions using BayesRC and imputed sequence data from 1000 Bull Genomes were 2% more accurate than with 800k data. We could demonstrate the method identified causal mutations in some cases. Further improvements will come from more accurate imputation of sequence variant genotypes...

ANALISIS SKENARIO KEGAGALAN SISTEM UNTUK MENENTUKAN PROBABILITAS KECELAKAAN PARAH AP1000

Directory of Open Access Journals (Sweden)

D.T. Sony Tjahyani

2014-03-01

Full Text Available Kejadian Fukushima telah menunjukkan bahwa kecelakaan parah dapat terjadi, maka dari itu sangatlah penting untuk menganalisis tingkat keselamatan pada reaktor daya. Berdasarkan rekomendasi expert mission IAEA setelah kejadian Fukushima, perlu dilakukan upaya untuk meminimalisasi terjadinya kecelakaan parah yaitu dengan melakukan proses pendinginan yang maksimal. Dalam konsep keselamatan fasilitas nuklir, khususnya reaktor daya telah diterapkan konsep keselamatan berlapis (Defence in Depth, DiD. Konsep keselamatan tersebut terdiri atas 5 level pertahanan yang bertujuan mencegah dan mengurangi lepasan produk fisi ke masyarakat dan lingkungan pada saat reaktor daya mengalami kecelakaan. Dalam reaktor telah didesain sistem atau tindakan yang mempunyai fungsi untuk mengatasi setiap level tersebut. Tujuan dari analisis ini adalah menentukan probabilitas kecelakaan parah dengan melakukan skenario kegagalan sistem dalam proses pendinginan di reaktor. Sebagai obyek analisis adalah reaktor daya AP1000, karena jenis reaktor ini sedang banyak dibangun saat ini. Skenario dilakukan dengan mengasumsikan beberapa kombinasi kegagalan sistem yang termasuk dalam DiD level 2 dan 3. Kegagalan sistem kemudian dianalisis dengan menggunakan analisis pohon kegagalan berdasarkan perangkat lunak SAPHIRE ver. 6.76. Dari analisis didapatkan probabilitas gagal dari kelompok sistem DiD level 2 dan 3 pada AP1000 masih di bawah batas kriteria dari IAEA yaitu lebih kecil dari 10-2, serta probabilitas kecelakaan parah didapatkan sebesar 6,17 x 10-10. Berdasarkan analisis ini disimpulkan bahwa AP1000 mempunyai tingkat keselamatan yang cukup tinggi, karena melalui skenario kegagalan sistem didapatkan probabilitas kecelakaan parah yang sangat kecil.   ABSTRACT Fukushima accident has shown that severe accident could be occurred, therefore it is important to analyze safety level of nuclear power plants. Based on the recommendations of IAEA expert mission after the Fukushima accident

77 FR 58829 - Notice of FERC Staff Attendance at Entergy/Cleco Teleconference on Order No. 1000

Science.gov (United States)

2012-09-24

... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission Notice of FERC Staff Attendance at Entergy/Cleco Teleconference on Order No. 1000 The Federal Energy Regulatory Commission (Commission.... 1000 Compliance September 19, 2012 (2-4pm EDT) The discussions may address matters at issue in the...

24 CFR 1000.118 - What recourse does a recipient have if HUD disapproves a proposal to provide assistance to non...

Science.gov (United States)

2010-04-01

... housing activity? 1000.118 Section 1000.118 Housing and Urban Development Regulations Relating to Housing... § 1000.118 What recourse does a recipient have if HUD disapproves a proposal to provide assistance to non...

24 CFR 1000.124 - What maximum and minimum rent or homebuyer payment can a recipient charge a low-income rental...

Science.gov (United States)

2010-04-01

... assisted with NAHASDA grant amounts? 1000.124 Section 1000.124 Housing and Urban Development Regulations... Activities § 1000.124 What maximum and minimum rent or homebuyer payment can a recipient charge a low-income...

25 CFR 1000.27 - How does the Director select which Tribes in the applicant pool become self-governance Tribes?

Science.gov (United States)

2010-04-01

... applicant pool become self-governance Tribes? 1000.27 Section 1000.27 Indians OFFICE OF THE ASSISTANT... for Participation in Tribal Self-Governance Admission into the Applicant Pool § 1000.27 How does the Director select which Tribes in the applicant pool become self-governance Tribes? The Director selects up...

24 CFR 1000.108 - How is HUD approval obtained by a recipient for housing for non low-income Indian families and...

Science.gov (United States)

2010-04-01

... recipient for housing for non low-income Indian families and model activities? 1000.108 Section 1000.108... AMERICAN HOUSING ACTIVITIES Affordable Housing Activities § 1000.108 How is HUD approval obtained by a.... Assistance to non low-income Indian families must be in accordance with § 1000.110. Proposals may be...

SWR 1000 related containment cooling system tests in PANDA

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

2000-01-01

Since 1991 the Paul Scherrer Institute has participated in the investigations of several of the new passive Advanced Light Water Reactor designs proposed world-wide. The current phase of the project, ALPHA-II, is focused on both the boiling water and the pressurized water reactor passive designs and consists of three projects under the sponsorship of the European Commission. The paper describes the performed PANDA transient system tests related to one of these projects, called 'BWR R and D Cluster for Innovative Passive Safety Systems (IPSS)', and details the PSI contribution to the experimental investigation of passive containment cooling by a Building Condenser system which is part of the advanced Boiling Water Reactor SWR 1000 designed by Siemens. First, a short description of the relevant systems of the SWR 1000 design and its simulation in the PANDA facility are presented. After the description of the experimental programme for the large-scale integral system test investigations in the PANDA facility, the main results of the performed tests are also given. Finally, the main conclusions, based on the to date available experimental results and their analysis, are summarised. (author)

Experience and prospects of WWER-1000 reactor spent fuel transport

International Nuclear Information System (INIS)

Kondratyev, A.N.; Yershov, V.N.; Kozlov, Yu.V.; Kosarev, Yu.A.; Ilyin, Yu.V.; Pavlov, M.S.

1989-01-01

The paper deals with the USSR experience in shipping the commercial WWER-1000 reactor spent fuel in TK-10 and TK-13 casks. The cask designs, their basic characteristics and the WWER-1000 spent fuel features are described. An example of calculational/experimental approach in the design of a basket (one of the most important components) for spent fuel assembly (SFA) accommodation in a cask is given. The main problems of future development works are presented in brief. A concept of development of nuclear power industry with the closed fuel cycle is assumed in the Soviet Union, hence the spent nuclear fuel is to be transported from NPPs to reprocessing plants. To transport WWER-1000 spent fuel, the casks of two types were developed. These are: a pilot TK-10 cask of 3t capacity in fuel; a commercial TK-13 cask of ∼6t capacity in fuel. The pilot TK-10 cask is thick-walled (360mm) cylindrical vessel manufactured of steel shells and a bottom welded to each other. The material of the body is carbon steel. There is a steel jacket on the outer side of the cask body and at 120 mm distance off the bottom. On its cylindrical part between the jacket and the body there are T-shaped circular ribs acting as shock-absorbers. The space between the jacket and the body is filled with ethylene glycol solution of 65 degree C crystallization temperature, which functions as a neutron shielding. The TK-10 cask coolant is water or air (nitrogen) at minor excess pressure resulted from FA heatup after the cask sealing

The Presto 1000: A novel automated transcranial Doppler ultrasound system.

Science.gov (United States)

Han, Seunggu J; Rutledge, William Caleb; Englot, Dario J; Winkler, Ethan A; Browne, Janet L; Pflugrath, Lauren; Cronsier, David; Abla, Adib A; Kliot, Michel; Lawton, Michael T

2015-11-01

We examined the reliability and ease of use of a novel automated transcranial Doppler (TCD) system in comparison to a conventional TCD system. TCD ultrasound allows non-invasive monitoring of cerebral blood flow, and can predict arterial vasospasm after a subarachnoid hemorrhage (SAH). The Presto 1000 TCD system (PhysioSonics, Bellevue, WA, USA) is designed for monitoring flow through the M1 segment of the middle cerebral artery (MCA) via temporal windows. The Presto 1000 system was tested across multiple preclinical and clinical settings in parallel with a control predicate TCD system. In a phantom flow generating device, both the Presto 1000 and Spencer system (Spencer Technologies, Redmond, WA, USA) were able to detect velocities with high accuracy. In nine volunteer patients, the Presto system was able to locate the MCA in 14 out of 18 temporal windows, in an average of 12.5s. In the SAH cohort of five patients with a total of 25 paired measurements, the mean absolute difference in flow velocities of the M1 segment, as measured by the two systems, was 17.5 cm/s. These data suggest that the Presto system offers an automated TCD that can reliably localize and detect flow of the MCA, with relative ease of use. The system carries the additional benefit of requiring minimal training for the operator, and can be used by many providers across multiple bedside settings. The mean velocities that were generated warrant further validation across an extended group of patients, and the predictive value for vasospasm should be checked against the current standard of angiography. Copyright © 2015 Elsevier Ltd. All rights reserved.

24 CFR 1000.36 - How long must a recipient retain records regarding exceptions made to the conflict of interest...

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false How long must a recipient retain records regarding exceptions made to the conflict of interest provisions? 1000.36 Section 1000.36 Housing... AMERICAN HOUSING ACTIVITIES General § 1000.36 How long must a recipient retain records regarding exceptions...

25 CFR 1000.229 - May a Tribe/Consortium propose a substitute for a regulation it wishes to be waived?

Science.gov (United States)

2010-04-01

... regulation it wishes to be waived? 1000.229 Section 1000.229 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Waiver of Regulations § 1000.229 May a Tribe...

Thermogravimetric analysis of silicon carbide-silicon nitride fibers at ambient to 1000 C in air

Science.gov (United States)

Daniels, J. G.; Ledbetter, F. E., III; Clemons, J. M.; Penn, B. G.

1984-01-01

Thermogravimetric analysis of silicon carbide-silicon nitride fibers was carried out at ambient to 1000 C in air. The weight loss over this temperature range was negligible. In addition, the oxidative stability at high temperature for a short period of time was determined. Fibers heated at 1000 C in air for fifteen minutes showed negligible weight loss (i.e., less than 1 percent).

Analysis of LOCV(Loss Of Condenser Vacuum) for OPR1000 using the SPACE

Energy Technology Data Exchange (ETDEWEB)

Youn, Bum-Soo; Kim, Se-Yun; Lee, Dong-Hyuk [KHNP CRI, Daejeon (Korea, Republic of)

2015-10-15

For validation of SPACE code developed by KHNP, the LOCV(loss of condenser vacuum) accident for OPR1000 is analyzed with SPACE code and results are compared with those from RETRAN code. The LOCV can occur from break in the circulation system, break in the condenser vacuum system, etc. The analysis assumes turbine trip at the beginning of the accident. In the actual plant, the steam bypass control system and the reactor power cutback system reduce the pressure increase of the pressurizer and steam generator from turbine trip. However, in this analysis, steam bypass system and reactor cutback system are not used. The LOCV result in stopping of steam flow to turbine and feed water flow to steam generator. This reduces heat removal to the secondary side and causes pressure increase in primary and secondary side. Using the SPACE, the thermal hydraulic behavior of the LOCV accident for OPR1000 was analyzed. The results are compared with RETRAN code. The results of two codes were very similar. From these results, SPACE has the ability to simulation LOCV for OPR1000.

A neuroprotective role for microRNA miR-1000 mediated by limiting glutamate excitotoxicity

DEFF Research Database (Denmark)

Verma, Pushpa; Augustine, George J; Ammar, Mohamed-Raafet

2015-01-01

Evidence has begun to emerge for microRNAs as regulators of synaptic signaling, specifically acting to control postsynaptic responsiveness during synaptic transmission. In this report, we provide evidence that Drosophila melanogaster miR-1000 acts presynaptically to regulate glutamate release at ...... a neuroprotective function in the brains of flies and mice. Drosophila miR-1000 showed activity-dependent expression, which might serve as a mechanism to allow neuronal activity to fine-tune the strength of excitatory synaptic transmission....

16 CFR 1000.20 - Office of Information and Public Affairs.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Office of Information and Public Affairs... ORGANIZATION AND FUNCTIONS § 1000.20 Office of Information and Public Affairs. The Office of Information and Public Affairs, which is managed by the Director of the Office, is responsible for the development...

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI Line Break

Directory of Open Access Journals (Sweden)

A.S. Ekariansyah

2016-08-01

Full Text Available Generation II Nuclear Power Plants (NPPs have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

Current status of generation III nuclear power and assessment of AP1000 developed by Westinghouse

International Nuclear Information System (INIS)

Zhang Mingchang

2005-01-01

In order to make greater contributions to the environment, new nuclear power systems will be needed to meet the increase of electricity demand and to replace plants to be decommissioned. A series of new designs, so called Generation III and Generation III +, are being developed to ensure their deployment in a Near-Term Deployment Road-map in US by 2010 and in Europe by 2015. The AP1000, developed by Westinghouse, is a two-loop 1000 MWe PWR with passive safety features and extensive simplifications to enhance its competitiveness in cost and tariff. It is the first Generation III + plant receiving the Final Design Approval by the US NRC. This paper briefly describes AP1000 design features and technical specifications, and presents a more detailed design evaluation with reference to relevant literatures. Both the opportunity and challenges for nuclear power development in China during the first decade of the 21 st century in a historic transition from Gen II to Gen III are analyzed. The key is to balance risks and benefits if the first AP1000 to be settled down in China. (author)

Spent fuel pool spray cooling system for the AP1000 {sup registered}

Energy Technology Data Exchange (ETDEWEB)

Vujic, Zoran; Sassen, Felix; Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2013-07-01

The AP1000 {sup registered} plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for Design Basis Events and Beyond Design Basis Accidents (BDBA). The AP1000 {sup registered} plant lines of defense with respect to Spent Fuel Pool (SFP) cooling are as follows: 1. During normal and abnormal conditions, defense-in-depth and duty systems provide highly reliable SFP cooling, supplied by offsite AC power or the onsite Standby Diesel Generators. 2. For unlikely events with extended loss of AC power (i.e. station black-out) and/or loss of heat sink, spent fuel cooling can be still provided indefinitely by: 2a. Passive systems, requiring minimal or no operator actions, sufficient for at least 72 hours under all possible loading conditions. 2b. After 3 days, several different means are provided to continue SFP cooling using installed plant equipment as well as off-site equipment with built-in connections. 3. Even for BDBA with postulated SFP damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 {sup registered} SFP Spray System provides an additional line of defense to prevent spent fuel damage. (orig.)

Analysis of small break loss of coolant accident for Chinese CPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

2016-05-15

This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

25 CFR 1000.45 - How many grants will the Department make each year and what funding will be available?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false How many grants will the Department make each year and what funding will be available? 1000.45 Section 1000.45 Indians OFFICE OF THE ASSISTANT SECRETARY... Availability, Amount, and Number of Grants § 1000.45 How many grants will the Department make each year and...

24 CFR 1000.34 - What factors must be considered in making an exception to the conflict of interest provisions?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What factors must be considered in making an exception to the conflict of interest provisions? 1000.34 Section 1000.34 Housing and Urban... ACTIVITIES General § 1000.34 What factors must be considered in making an exception to the conflict of...

75 FR 70861 - Airworthiness Directives; Fokker Services B.V. Model F.28 Mark 0100, 1000, 2000, 3000, and 4000...

Science.gov (United States)

2010-11-19

....V. Model F.28 Mark 0100, 1000, 2000, 3000, and 4000 Airplanes AGENCY: Federal Aviation... Fokker Services B.V. Model F28 Mark 1000, 2000, 3000, and 4000 airplanes, all serial numbers, equipped... Instructions of Fokker Service Bulletin SBF28-28-052, dated April 20, 2010 (for Model F28 Mark 1000, 2000, 3000...

Fusion of eastern and western technology in VVER 1000 NPP upgrade

International Nuclear Information System (INIS)

Ubra, O.; Fleischhans, J.; Kveton, M.

1997-01-01

An extensive modernization program upgrading two units of VVER 1000 type of the Czech nuclear power plant (NPP) Temelin to meet the latest international standards is presented. The program is based primarily on combination of eastern and western technology and it has been implemented during plant construction. The NPP Temelin was originally designed according to the standards of the former Soviet Union. After a series of reviews in the 1990s, a decision was made by the Temelin management of upgrade the design of the plant, including the supply of fuel and instrumentation and control system by a western company. The adoption of western technology and practices has helped to solve a large number of IAEA safety issues related to design and operation of VVER 1000 NPP. Details on the current Temelin design and other related safety matters are presented

Research on alternative measures in the F1000 system with Google Scholar citation index

Directory of Open Access Journals (Sweden)

Saeideh Ebrahimy

2016-08-01

Full Text Available The purpose of this study was to assess the relationship between altmetrics measures of F1000 website and citation Indexes of Google scholar. Applied research and survey methods, citation analysis, and webometrics with descriptive analytical approach are used. The study population comprised 218 biomedical research papers have been indexed in F1000 system in the period 2012-2014. A sample of 100 research articles were purposely selected from F1000 system based on the purpose of the research. Variables associated with altmetrics measures were extracted from F1000 and variables associated with citation measures were from Google scholar. Data analysis was conducted by SPSS software version 16 using descriptive and inferential statistics. The results of this study indicate a significant positive correlation between variables associated with altmetrics and citation measures. They also confirm a strong positive correlation between Altmetrics indicators. However, the measures of the number of the authors and the funding supply of the articles show no significant correlation with altmetrics and citation measures. So an Altmetrics measure presents a new way to measure the impact of writers and publications which is a complementary to traditional assessment indicators. The significant correlation between altmetrics and citation metrics indicate that combination of traditional and altmetrics indicators presents a more complete history of the writer or the article  which provides clear dimensions of scientific assessment of research.

The comparison of System 1000 analysis and type testing for neoprene gasket environmental qualification

International Nuclear Information System (INIS)

Park, Kyung Heum; Kim, Jong Seok; Jeong, Sun Chul; Jang, Kyung Nam; Hwang, Sung Phil

2010-01-01

The typical environmental qualification is to ensure that equipment will operate on demand to meet system performance requirements during normal and abnormal service conditions. There are four environmental qualification methods, type testing, operating experience, analysis and combined method. Generally, the American EQ do not contain the mechanical equipment like pumps and valves in their EQ equipment list because their EQ standard 10CFR50.49 limits EQ equipment as electrical equipment. On the other hand, Canadian EQ contain the mechanical equipment like pumps and valves in their EQ components list, Canadians usually call American 'equipment' as 'components', because their EQ standard CSA N290.13-05 do not limits EQ equipment as electrical equipment. System 1000 program is typical Canadian EQ analysis method using mathematical modeling and comparison with established engineering information and manufacturers' data. Most of Canadian nuclear power utilities like NB Power, Hydro Quebec and OPG use the System 1000 program to evaluate the design life for their EQ components. To qualify a pump, I had to list all the non-metallic parts in the pump and found there are lots of gaskets made by neoprene material. I tried to qualify these neoprene gaskets by analysis using System 1000 program and by type testing. In this paper, I'd like to introduce the qualification results of neoprene gasket both type testing and analysis using System 1000 program

25 CFR 1000.398 - May a Tribe/Consortium invest funds received under a self-governance agreement?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false May a Tribe/Consortium invest funds received under a self-governance agreement? 1000.398 Section 1000.398 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS... invest funds received under a self-governance agreement? Yes, self-governance funds may be invested if...

Predisposal of Radioactive Waste from NPP 1000 MWe

International Nuclear Information System (INIS)

Suryantoro

2007-01-01

Predisposal of radioactive waste from NPP 1000 MW which was planned to be operated in 2016 has been conducted. In this study NPP applying PWR type was assumed. This assessment comprises all aspects of radioactive waste coming from NPP. One through cycle was chosen consequently no reprocessing step will be conducted. The assessment shows that technologically all radioactive waste treatment process rising from NPP operation has similarities to the existing radioactive waste process conducted by RWI which has lower scale of waste amount. (author)

Addressing the characterisation challenge to understand catalysis in MOFs: the case of nanoscale Cu supported in NU-1000.

Science.gov (United States)

Platero-Prats, Ana E; Li, Zhanyong; Gallington, Leighanne C; Peters, Aaron W; Hupp, Joseph T; Farha, Omar K; Chapman, Karena W

2017-09-01

We explore the dynamic structure and reactivity of Cu species supported on NU-1000. By combining pair distribution function (PDF) analysis and difference envelope density (DED) analysis of in situ synchrotron-based X-ray scattering data, we simultaneously probe the local structure of supported Cu-species, their distribution within NU-1000 and distortions of the NU-1000 lattice under conditions relevant to catalysis and catalyst activation. These analyses show that atomic layer deposition (ALD) of Cu in NU-1000 (Cu-AIM) leads to the formation of Cu-oxo clusters within the small pores that connect the triangular and hexagonal channels. Exposure of Cu-AIM to a reducing atmosphere at 200 °C produces metallic Cu 0 of two distinct particle sizes: ∼4 nm nanoparticles and small sub-nanometer clusters. The size of these nanoparticles appears to be constrained by NU-1000 pore dimensions, with evidence of the sub-nanometer clusters being bound within the triangular channels flanked by pyrene rings. This supported Cu 0 -NU-1000 system is catalytically active for gas-phase ethylene hydrogenation. Exposure of the catalyst to oxidative atmosphere re-oxidises the Cu species to a Cu 2 O cuprite phase. The dynamic restructuring of the system in different chemical environments underscores the importance of probing these systems in situ.

76 FR 6541 - Airworthiness Directives; Fokker Services B.V. Model F.28 Mark 0100, 1000, 2000, 3000, and 4000...

Science.gov (United States)

2011-02-07

... Airworthiness Directives; Fokker Services B.V. Model F.28 Mark 0100, 1000, 2000, 3000, and 4000 Airplanes AGENCY.... Applicability (c) This AD applies to Fokker Services B.V. Model F.28 Mark 1000, 2000, 3000, and 4000 airplanes... April 20, 2010 (for Model F.28 Mark 1000, 2000, 3000, and 4000 airplanes); or SBF100-28-063, dated April...

25 CFR 1000.73 - Once a Tribe/Consortium has been awarded a grant, may the Tribe/Consortium obtain information...

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false Once a Tribe/Consortium has been awarded a grant, may the Tribe/Consortium obtain information from a non-BIA bureau? 1000.73 Section 1000.73 Indians OFFICE OF THE... § 1000.73 Once a Tribe/Consortium has been awarded a grant, may the Tribe/Consortium obtain information...

24 CFR 1000.116 - What should HUD do before declining a proposal to provide assistance to non low-income Indian...

Science.gov (United States)

2010-04-01

... a proposal to provide assistance to non low-income Indian families or a model housing activity? 1000.116 Section 1000.116 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES Affordable Housing Activities § 1000.116 What...

SWR 1000: A Next-Generation Boiling Water Reactor Ready for Deployment

International Nuclear Information System (INIS)

Brettschuh, W.

2006-01-01

The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 design is particularly suitable for countries whose power systems do not include any large power plants. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (IandC) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10x10 rod array to a 12x12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free start-up, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving

The estimation of the control rods absorber burn-up during the VVER-1000 operation

Energy Technology Data Exchange (ETDEWEB)

Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2013-09-15

The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

Study on time management of FCD concrete pouring AP1000 NPP and its application

International Nuclear Information System (INIS)

Wu Jie; Yang Ming; Cong Jiuyuan

2010-01-01

Haiyang nuclear power plant adopts the advanced third-generation nuclear power technology, AP1000, the design concept of passive system and the methods of modularization and 'open-top' construction greatly shortens the construction period. At the same time, higher requirements are put forward to the management of construction, quality and progress control. This paper will apply the statistical process control theory and method to the time management of FCD (First Concrete Day) to Unit 1 of Haiyang AP1000 nuclear power plant, and it brings up the mathematical model of time management based on verification through modeling, data analysis, model optimization and the actual construction work. The theory and method studied in this paper can not be only applied to the FCD concrete pouring for Unit 1 of Haiyang AP1000 nuclear power plant, but also have great referential and guiding significance to the continuous concreting of the mass concrete of the follow-up similar construction. (authors)

Natural circulation and stratification in the various passive safety systems of the SWR 1000

International Nuclear Information System (INIS)

Meseth, J.

2002-01-01

In some of the passive safety systems of Siemens' SWR 1000 boiling water reactor (i.e. the emergency condensers and containment cooling condensers), natural circulation is the main effect on both the primary and secondary sides by which optimum system efficiency is achieved. Other passive safety systems of the SWR 1000 require natural circulation on the secondary side only (condensation of steam discharged by the safety and relief valves; cooling of the Reactor Pressure Vessel (RPV) by flooding from the outside in case of core melt), while still other systems require stratification to be effective (i.e. the passive pressure pulse transmitters and steam-driven scram tanks). Complex natural circulation and stratification can take place simultaneously if fluids with different densities are enclosed in a single volume (in a core melt accident, for example, the nitrogen, steam and hydrogen in the containment). Related problems and the solutions thereto planned for the SWR 1000 are reported from the designer's viewpoint. (author)

AP1000 will meet the challenges of near-term deployment

International Nuclear Information System (INIS)

Matzie, Regis A.

2008-01-01

The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not have access to electricity or clean water. Climate change and the concern for increased emissions of green house gases have brought into question the future primary reliance of fossil fuels. With the projected worldwide increase in energy demand, concern for the environmental impact of carbon emissions, and the recent price volatility of fossil fuels, nuclear energy is undergoing a rapid resurgence. This 'nuclear renaissance' is broad based, reaching across Asia, North America, Europe, as well as selected countries in Africa and South America. Many countries have publicly expressed their intentions to pursue the construction of new nuclear energy plants. Some countries that have previously turned away from commercial nuclear energy are reconsidering the advisability of this decision. This renaissance is facilitated by the availability of more advanced reactor designs than are operating today, with improved safety, economy, and operations. One such design, the Westinghouse AP1000 advanced passive plant, has been a long time in the making! The development of this passive technology started over two decades ago from an embryonic belief that a new approach to design was needed to spawn a nuclear renaissance. The principal challenges were seen as ensuring reactor safety by requiring less reliance on operator actions and overcoming the high plant capital cost of nuclear energy. The AP1000 design is based on the use of innovative passive technology and modular construction, which require significantly less equipment and commodities that facilitate a more rapid construction schedule. Because Westinghouse had the vision and the perseverance to continue the development of this passive technology, the AP1000 design is ready to meet today's challenge of near-term deployment

1,000 Days: Mobilizing Investments for Healthier, More Prosperous Futures

Science.gov (United States)

Sullivan, Lucy Martinez; Sakayan, Mannik; Cernak, Kimberly

2018-01-01

Good nutrition during the 1,000-day window between pregnancy and 2 years old can give children the opportunity to reach their full potential. Conversely, malnutrition early in life can cause irreversible damage to a child's brain development and physical growth, leading to a lifetime of poor health and lost potential. Each year, malnutrition costs…

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

Methane Oxidation to Methanol Catalyzed by Cu-Oxo Clusters Stabilized in NU-1000 Metal-Organic Framework.

Science.gov (United States)

Ikuno, Takaaki; Zheng, Jian; Vjunov, Aleksei; Sanchez-Sanchez, Maricruz; Ortuño, Manuel A; Pahls, Dale R; Fulton, John L; Camaioni, Donald M; Li, Zhanyong; Ray, Debmalya; Mehdi, B Layla; Browning, Nigel D; Farha, Omar K; Hupp, Joseph T; Cramer, Christopher J; Gagliardi, Laura; Lercher, Johannes A

2017-08-02

Copper oxide clusters synthesized via atomic layer deposition on the nodes of the metal-organic framework (MOF) NU-1000 are active for oxidation of methane to methanol under mild reaction conditions. Analysis of chemical reactivity, in situ X-ray absorption spectroscopy, and density functional theory calculations are used to determine structure/activity relations in the Cu-NU-1000 catalytic system. The Cu-loaded MOF contained Cu-oxo clusters of a few Cu atoms. The Cu was present under ambient conditions as a mixture of ∼15% Cu + and ∼85% Cu 2+ . The oxidation of methane on Cu-NU-1000 was accompanied by the reduction of 9% of the Cu in the catalyst from Cu 2+ to Cu + . The products, methanol, dimethyl ether, and CO 2 , were desorbed with the passage of 10% water/He at 135 °C, giving a carbon selectivity for methane to methanol of 45-60%. Cu oxo clusters stabilized in NU-1000 provide an active, first generation MOF-based, selective methane oxidation catalyst.

1000 kW ICRH amplifiers for MFTF-B

International Nuclear Information System (INIS)

Boksberger, U.

1986-01-01

For the startup of the MFTF-B ICRH heating will be applied. Two commercial amplifiers derived from standard broadcast transmitters provide 1000 kW RF power each into a matching system for any VSWR as high as 1.5. Emphasis is put on the specific environment of magnetic fields and seismic loads as well as to the particular RF power control requirements and remote operation. Also addressed is the amplifier's performance into a typical load. The load variations due to the MFTF-B plasma coupling were calculated by TRW

Proteomic biomarker discovery in 1000 human plasma samples with mass spectrometry

DEFF Research Database (Denmark)

Cominetti, Ornella; Núñez Galindo, Antonio; Corthésy, John

2016-01-01

automated proteomic biomarker discovery workflow. Herein, we have applied this approach to analyze 1000 plasma samples from the multicentered human dietary intervention study "DiOGenes". Study design, sample randomization, tracking, and logistics were the foundations of our large-scale study. We checked...

VVER-1000 small-medium break LOCAs predictions by ASTEC

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Atanasova, B.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

2005-01-01

This paper deals with an assessment of ASTEC1.1v0 code in the simulation of small and medium break LOCAs (ranging from 30mm up to 70mm equivalent diameter). The reference power plant for this analysis is a VVER-1000/V320 (e.g. Units 5 and 6 at Kozloduy NPP). A preliminary comparison with MELCOR and RELAP-SCDAP severe accident codes will be discussed. This investigation has been performed in the framework of the SARNET project (under the Euratom 6 th framework program) by the FoBAUs group (Forum of Bulgarian ASTEC users). The FoBAUs group aims at the validation of the ASTEC code in the field of severe accidents. Future activities will target the ASTEC capability (as a PSA-level 2 tool) to simulate a large range of reactor accident scenarios with intervention of safety systems (either passive systems or operated by operators). The final target is to assess Severe Accident Management (SAM) procedures for VVER-1000 reactors. The ASTEC1.1v0 code version here used is the one released in June 2004 by the French IRSN (Institut de Radioprotection et de Surete Nucleaire) and the German GRS (Gesellschaft ReactorSicherheit mbH). (author)

EP1000 anticipated transient without scram analyses

International Nuclear Information System (INIS)

Saiu, G.; Frogheri, M.; Schulz, T.L.

2001-01-01

The present paper summarizes the main results of the Anticipated Transient Without Scram (ATWS) analysis activity, performed for the European Passive Plant Program (EPP). The behavior of the EP1000 plant following an ATWS has been analyzed by means of the RELAP5/Mod3.2 code. An ATWS is defined as an Anticipated Transient accompanied by a common mode failure in the reactor protection system, such that the control rods do not scram as required to mitigate the consequences of the transient. According to the experience gained in PWR design, the limiting ATWS events, in a PWR, have been found to be the heatup transients caused by a reduction of heat removal capability by the secondary side of the plant. For this reason, the Loss of Normal Feedwater initiating event, to which the failure of the reactor scram is associated, has been analyzed. The purpose of the study is to verify the performance requirements set for the core feedback characteristics (that is to evaluate the effect of the low boron core neutron kinetic parameters), the overpressure protection system, and boration systems to cope with the EUR Acceptance Criteria for ATWS. Another purpose of this analysis was to support development of revised PSA success criteria that would reduce the contribution of ATWS to the large release frequency (LRF). The low boron core improved the basic EP1000 response to an ATWS event. In particular, the peak pressure was significantly lower than that which would result from a standard core configuration. The improved ATWS analysis results also permitted improved ATWS PSA success criteria. For example, the reduced peak pressure allows the use of other plant features to mitigate the event, including manual initiation of feed-bleed cooling in the event of PRHR HX failure. As a result, the core melt frequency and especially the LRF are significantly reduced. (author)

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

Confirmatory analysis of the AP1000 passive residual heat removal heat exchanger with 3-D computational fluid dynamic analysis

International Nuclear Information System (INIS)

Schwall, James R.; Karim, Naeem U.; Thakkar, Jivan G.; Taylor, Creed; Schulz, Terry; Wright, Richard F.

2006-01-01

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power up-rate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model

Low cycle fatigue behavior of polycrystalline NiAl at 300 and 1000 K

Science.gov (United States)

Lerch, Bradley A.; Noebe, Ronald D.

1993-01-01

The low cycle fatigue behavior of polycrystalline NiAl was determined at 300 and 1000 K - temperatures below and above the brittle- to-ductile transition temperature (BDTT). Fully reversed, plastic strain-controlled fatigue tests were conducted on two differently fabricated alloy samples: hot isostatically pressed (HIP'ed) prealloyed powder and hot extruded castings. HIP'ed powder (HP) samples were tested only at 1000 K, whereas the more ductile cast-and-extruded (C+E) NiAl samples were tested at both 1000 and 300 K. Plastic strain ranges of 0.06 to 0.2 percent were used. The C+E NiAl cyclically hardened until fracture, reaching stress levels approximately 60 percent greater than the ultimate tensile strength of the alloy. Compared on a strain basis, NiAl had a much longer fatigue life than other B2 ordered compounds in which fracture initiated at processing-related defects. These defects controlled fatigue life at 300 K, with fracture occurring rapidly once a critical stress level was reached. At 1000 K, above the BDTT, both the C+E and HP samples cyclically softened during most of the fatigue tests in air and were insensitive to processing defects. The processing method did not have a major effect on fatigue life; the lives of the HP samples were about a factor of three shorter than the C+E NiAl, but this was attributed to the lower stress response of the C+E material. The C+E NiAl underwent dynamic grain growth, whereas the HP material maintained a constant grain size during testing. In both materials, fatigue life was controlled by intergranular cavitation and creep processes, which led to fatigue crack growth that was primarily intergranular in nature. Final fracture by overload was transgranular in nature. Also, HP samples tested in vacuum had a life three times longer than their counterparts tested in air and, in contrast to those tested in air, hardened continuously over half of the sample life, thereby indicating an environmentally assisted fatigue damage

Feasibility analysis of aggressive cooldown in OPR-1000 nuclear power plants

International Nuclear Information System (INIS)

Kim, Man Cheol; Jerng, Dong Wook

2014-01-01

Highlights: • We examine the feasibility of aggressive cooldown, particularly in OPR-1000 plants. • We review experimental and simulation results, EOPs, and time estimates. • Aggressive cooldown is, in general, expected to be performed successfully. • There is potential for failure to initiate aggressive cooldown in a timely manner. • Investigation on allowing higher cooldown rate in OPR-1000 plants is recommended. - Abstract: Aggressive cooldown is the action taken by main control room (MCR) operators to rapidly cool down and depressurize the reactor coolant system at the maximum allowed cooldown rate during a small break loss of coolant accident (SBLOCA) or a steam generator tube rupture accident with failure of high-pressure safety injection (HPSI) in a pressurized water reactor. We examined the feasibility of aggressive cooldown in OPR-1000 nuclear power plants by establishing success criteria based on experimental results and thermal–hydraulic code simulations, reviewing emergency operating procedures, and estimating the necessary time for MCR operators to initiate aggressive cooldown based on experimental results with licensed nuclear power plant operators. For an SBLOCA with a break size of 0.02 ft 2 combined with total failure of HPSI, we found that aggressive cooldown can generally be performed successfully, but there is potential for failure to initiate aggressive cooldown in a timely manner. We discuss the potential effects of introducing the Ultimate Procedure on reducing the core damage potential. Detailed analysis and further discussion are necessary to reduce uncertainties associated with aggressive cooldown and its positive effect on the safety of nuclear power plants

25 CFR 1000.228 - What are examples of waivers prohibited by law?

Science.gov (United States)

2010-04-01

... ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN SELF-DETERMINATION AND EDUCATION ACT Waiver of Regulations § 1000.228 What are examples of waivers prohibited by law...

Strength analysis of PGV-1000M steam generator support

International Nuclear Information System (INIS)

Dubik, Ya.R.; Ageev, S.M.; Orynyak, I.V.; Vasilchenko, B.M.

2017-01-01

The paper presents the design of PGV-1000M steam generator support. It is shown that the load in the rolling support is distributed extremely unevenly, which is associated with the compliance of the support construction. It is demonstrated that under working loads only several rollers are used, the stresses in which exceed the yield strength. This can be an additional loading factor to be considered in the analysis of welding No. 111 failure.

Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

Energy Technology Data Exchange (ETDEWEB)

Hashim, Muhammad, E-mail: hashimsajid@yahoo.com; Hidekazu, Yoshikawa, E-mail: yosikawa@kib.biglobe.ne.jp; Takeshi, Matsuoka, E-mail: mats@cc.utsunomiya-u.ac.jp; Ming, Yang, E-mail: myang.heu@gmail.com

2014-10-15

Highlights: • Discussion on reasons why AP1000 equipped with ADS system comparatively to PWR. • Clarification of full and partial depressurization of reactor coolant system by ADS system. • Application case study of four stages ADS system for reliability evaluation in LBLOCA. • GO-FLOW tool is capable to evaluate dynamic reliability of passive safety systems. • Calculated ADS reliability result significantly increased dynamic reliability of PXS. - Abstract: AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs. In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated

Investigations in the upper load range of stratified operation with injection pressures up to 1000 bar; Untersuchungen im oberen Lastbereich des Schichtbetriebes bei Einspritzdruecken bis 1000 bar

Energy Technology Data Exchange (ETDEWEB)

Buri, Stefan; Busch, Steve; Kubach, Heiko; Spicher, Ulrich [Karlsruhe Univ. (DE). Inst. fuer Kolbenmaschinen (IFKM)

2010-07-01

This paper presents the results of recent research that has been performed on a single-cylinder spray-guided DISI engine at the Institut fuer Kolbenmaschinen. A production multihole injector is used as a baseline and compared with a specially adapted injector at higher injection pressures. Injection pressures of up to 1000 bar are utilized to investigate the combustion and emissions characteristics at the stability limit of stratified combustion. With the modified injector, measurements with the two-color method are applied to analyze this operating condition in terms of soot formation. Included are results for a spark timing variation with 1000 bar injection pressure, an injection pressure variation with constant spark timing, and fuel consumption-optimized engine operating parameters in order to analyze realistic operating points. The benefits of injecting fuel at higher pressures are described, as are the limitations of the experimental setup. (orig.)

16 CFR 1000.19 - Office of Financial Management, Planning and Evaluation.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Office of Financial Management, Planning and... COMMISSION ORGANIZATION AND FUNCTIONS § 1000.19 Office of Financial Management, Planning and Evaluation. The Office of Financial Management, Planning and Evaluation is responsible for developing the Commission's...

24 CFR 1000.236 - What are eligible administrative and planning expenses?

Science.gov (United States)

2010-04-01

... administrative management; (2) Coordination monitoring and evaluation; (3) Preparation of the IHP including data... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What are eligible administrative...) § 1000.236 What are eligible administrative and planning expenses? (a) Eligible administrative and...

Addressing the fundamental issues in reliability evaluation of passive safety of AP1000 for a comparison with active safety of PWR

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Yang Ming

2013-01-01

Passive safety systems adopted in advanced Pressurized Water Reactor (PWR), such as AP1000 and EPR, should attain higher reliability than the existing active safety systems of the conventional PWR. The objective of this study is to discuss the fundamental issues relating to the reliability evaluation of AP1000 passive safety systems for a comparison with the active safety systems of conventional PWR, based on several aspects. First, comparisons between conventional PWR and AP1000 are made from the both aspects of safety design and cost reduction. The main differences between these PWR plants exist in the configurations of safety systems: AP1000 employs the passive safety system while reducing the number of active systems. Second, the safety of AP1000 is discussed from the aspect of severe accident prevention in the event of large break loss of coolant accidents (LOCA). Third, detailed fundamental issues on reliability evaluation of AP1000 passive safety systems are discussed qualitatively by using single loop models of safety systems of both PWRs plants. Lastly, methodology to conduct quantitative estimation of dynamic reliability for AP1000 passive safety systems in LOCA condition is discussed, in order to evaluate the reliability of AP1000 in future by a success-path-based reliability analysis method (i.e., GO-FLOW). (author)

Parallel computing of a climate model on the dawn 1000 by domain decomposition method

Science.gov (United States)

Bi, Xunqiang

1997-12-01

In this paper the parallel computing of a grid-point nine-level atmospheric general circulation model on the Dawn 1000 is introduced. The model was developed by the Institute of Atmospheric Physics (IAP), Chinese Academy of Sciences (CAS). The Dawn 1000 is a MIMD massive parallel computer made by National Research Center for Intelligent Computer (NCIC), CAS. A two-dimensional domain decomposition method is adopted to perform the parallel computing. The potential ways to increase the speed-up ratio and exploit more resources of future massively parallel supercomputation are also discussed.

Systems required during and after an earthquake. Summary report. WWER-1000 nuclear power plants

International Nuclear Information System (INIS)

Monette, P.

1995-01-01

The scope of this document is to list the mechanical, instrumentation and electrical components required during and after earthquake, in order to achieve and maintain safe shutdown conditions of a WWER-1000 type nuclear power plant. The main objective pursued in establishing the systems and equipment list is to provide guidance for the design and implementation of the backfits which are necessary to increase seismic resistance of the components required after earthquake. The presented list is established on generic basis, i.e. it is applicable to any specific WWER-1000

78 FR 76813 - Igor Bobel, Inmate #-67253-066, FCI Loretto, Federal Correctional Institution, P.O. Box 1000...

Science.gov (United States)

2013-12-19

... DEPARTMENT OF COMMERCE Bureau of Industry and Security Igor Bobel, Inmate -67253-066, FCI Loretto, Federal Correctional Institution, P.O. Box 1000, Loretto, PA 15940; Order Denying Export Privileges On May... 1000, Loretto, PA 15940, and when acting for or on behalf of Bobel, his representatives, assigns...

25 CFR 1000.316 - May the Tribe/Consortium be reimbursed for actual and reasonable “wind up costs” incurred after...

Science.gov (United States)

2010-04-01

... reasonable âwind up costsâ incurred after the effective date of retrocession? 1000.316 Section 1000.316... Reassumption § 1000.316 May the Tribe/Consortium be reimbursed for actual and reasonable “wind up costs” incurred after the effective date of retrocession? Yes, the Tribe/Consortium may be reimbursed for actual...

Steam generator collector integrity of WWER-1000 reactors. IAEA extrabudgetary programme on the safety of WWER NPPs

Energy Technology Data Exchange (ETDEWEB)

Lin, C.; Strupczewski, A. [International Atomic Energy Agency, Vienna (Austria)

1995-12-31

At the Consultants` Meeting on `The Safety of WWER-1000 Model 320 Nuclear Power Plants` organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of WWER-1000 steam generator integrity was identified as an important issue of safety concern. Considering the safety importance of this issue, a Consultants` Meeting on `The Steam Generator Integrity of WWER-1000 Nuclear Power Plants` was convened in Vienna in May 1993, attended by 15 international experts in the area to compile information on the steam generator operating experience, deficiencies and corrective measures implemented and planned. In order to also include information from the main designer OKB Gidropress and to finalize the meeting report the IAEA convened a second meeting on the issue on 23-27 November 1993. The present paper summarizes the information and conclusions from those meetings.

Steam generator collector integrity of WWER-1000 reactors. IAEA extrabudgetary programme on the safety of WWER NPPs

Energy Technology Data Exchange (ETDEWEB)

Lin, C; Strupczewski, A [International Atomic Energy Agency, Vienna (Austria)

1996-12-31

At the Consultants` Meeting on `The Safety of WWER-1000 Model 320 Nuclear Power Plants` organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of WWER-1000 steam generator integrity was identified as an important issue of safety concern. Considering the safety importance of this issue, a Consultants` Meeting on `The Steam Generator Integrity of WWER-1000 Nuclear Power Plants` was convened in Vienna in May 1993, attended by 15 international experts in the area to compile information on the steam generator operating experience, deficiencies and corrective measures implemented and planned. In order to also include information from the main designer OKB Gidropress and to finalize the meeting report the IAEA convened a second meeting on the issue on 23-27 November 1993. The present paper summarizes the information and conclusions from those meetings.

Simulation of advanced accumulator and its application in CPR1000 LBLOCA analysis

International Nuclear Information System (INIS)

Hu, Hongwei; Shan, Jianqiang; Gou, Junli; Cao, Jianhua; Shen, Yonggang; Fu, Xiangang

2014-01-01

Highlights: • The analysis model was developed for advanced accumulator. • The sensitivity analysis of each key parameter was performed. • The LBLOCA was analyzed for the CPR1000 with advanced accumulator. • The analysis shows that advanced accumulator can improve CPR1000 safety performance. - Abstract: The advanced accumulator is designed to improve the safety and reliability of CPR1000 by providing a small injection flow to keep the reactor core in flooded condition. Thus, the core still stays in a cooling state without the intervention of low pressure safety injection and the startup grace time of the low pressure safety injection pump can be greatly extended. A new model for the advanced accumulator, which is based on the basic conservation equations, is developed and incorporated into RELAP5/MOD 3.3. The simulation of the advanced accumulator can be carried out and results show that the behavior of the advanced accumulator satisfied its primary design target. There is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, which results in a large pressure drop and a small flow. And then the sensitivity analysis is performed and the major factors which affected the flow rate of the advanced accumulator were obtained, including the damper diameter, the initial volume ratio of the water and the nitrogen and the diameter ratio of the standpipe and the small pipe. Additionally, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 with advanced accumulator is analyzed. The results show that the criterion for maximum cladding temperature limit (1477 K) (NRC, 1992) can be met ever with 200 s after the startup of the low pressure safety injection. From this point of view, passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, safety and economy of the reactor system can be improved

Pengaruh Edukasi Berbasis Keluarga terhadap Intensi Ibu Hamil untuk Optimalisasi Nutrisi pada 1000 Hari Pertama Kehidupan

Directory of Open Access Journals (Sweden)

Rosani Naim

2017-08-01

The findings showed that there was a significant effect between the family-based education towards pregnant mothers’ intention to optimize the nutrition of 1000 first days of life (p = 0.00. Thus, it is recommended to apply family based education in nursing intervention to improve of mothers and fetus health. Keywords : Family based education, intention, pregnant mothers’, nutrition, 1000 first days of life.

24 CFR 1000.130 - May a recipient charge a non low-income family rents or homebuyer payments which are more than 30...

Science.gov (United States)

2010-04-01

...? 1000.130 Section 1000.130 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES Affordable Housing Activities § 1000.130 May a...

24 CFR 1000.54 - What procedures apply to complaints arising out of any of the methods of providing for Indian...

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false What procedures apply to complaints arising out of any of the methods of providing for Indian preference? 1000.54 Section 1000.54 Housing and... ACTIVITIES General § 1000.54 What procedures apply to complaints arising out of any of the methods of...

24 CFR 1000.42 - Are the requirements of section 3 of the Housing and Urban Development Act of 1968 applicable?

Science.gov (United States)

2010-04-01

... 24 Housing and Urban Development 4 2010-04-01 2010-04-01 false Are the requirements of section 3 of the Housing and Urban Development Act of 1968 applicable? 1000.42 Section 1000.42 Housing and... ACTIVITIES General § 1000.42 Are the requirements of section 3 of the Housing and Urban Development Act of...

AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

Energy Technology Data Exchange (ETDEWEB)

Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel

1000kW on-site PAFC power plant development and demonstration

Energy Technology Data Exchange (ETDEWEB)

Satomi, Tomohide; Koike, Shunichi [Phosphoric Acid Fuel Cell Technology Research Association (PAFC-TRA), Osaka (Japan); Ishikawa, Ryou [New Energy and Industrial Technology Development Organization (NEDO), Tokyo (Japan)

1996-12-31

Phosphoric Acid Fuel Cell Technology Research Association (PAFC-TRA) and New Energy and Industrial Technology Development Organization (NEDO) have been conducting a joint project on development of a 5000kW urban energy center type PAFC power plant (pressurized) and a 1000kW on-site PAFC power plant (non-pressurized). The objective of the technical development of 1000kW on-site PAFC power plant is to realize a medium size power plant with an overall efficiency of over 70% and an electrical efficiency of over 36%, that could be installed in a large building as a cogeneration system. The components and system integration development work and the plant design were performed in 1991 and 1992. Manufacturing of the plant and installation at the test site were completed in 1994. PAC test was carried out in 1994, and generation test was started in January 1995. Demonstration test is scheduled for 1995 and 1996.

Characteristics and design improvement of AP1000 automatic depressurization system

International Nuclear Information System (INIS)

Jin Fei

2012-01-01

Automatic depressurization system, as a specialty of AP1000 Design, enhances capability of mitigating design basis accidents for plant. Advancement of the system is discussed by comparing with traditional PWR design and analyzing system functions, such as depressurizing and venting. System design improvement during China Project performance is also described. At the end, suggestions for the system in China Project are listed. (author)

Modernization incore monitoring system of WWER-1000 reactors (V-320)

International Nuclear Information System (INIS)

Mitin, Valentin; Semchenkov, Yurij; Kalinushkin, Andrey

2008-01-01

Modern ICIS system for VVER-1000, including a number of sensors, cable runs, corresponding measuring equipment and computer engineering, software, accumulated 30 year experience of interaction researches on VVER reactors and is capable to ensure carrying out of control, protection, informational, diagnostic functions and thus to promote real increase of quality, reliability and safety in nuclear fuel and NPP power units operation

Thermal-hydraulic characteristic of the PGV-1000 steam generator

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

AP1000 - update on projects in US and China

Energy Technology Data Exchange (ETDEWEB)

Godfrey, M. [Westinghouse Electric Company, Cranberry Township, Pennsy lvania (United States)

2012-07-01

Westinghouse is the only company solely focused on commercial nuclear technology. Westinghouse business is based on four product lines regionally divided: nuclear power plants, nuclear fuel, nuclear services and nuclear automation. The AP1000 is the technology of choice for more than half of the new plants identified in the US. Westinghouse has the only certified Generation III+ technology by the US Nuclear Regulatory Commission (NRC). The first Generation III+ plants are under construction in China and the US.

AP1000 - update on projects in US and China

International Nuclear Information System (INIS)

Godfrey, M.

2012-01-01

Westinghouse is the only company solely focused on commercial nuclear technology. Westinghouse business is based on four product lines regionally divided: nuclear power plants, nuclear fuel, nuclear services and nuclear automation. The AP1000 is the technology of choice for more than half of the new plants identified in the US. Westinghouse has the only certified Generation III+ technology by the US Nuclear Regulatory Commission (NRC). The first Generation III+ plants are under construction in China and the US.

Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

International Nuclear Information System (INIS)

Belousov, S.; Antonov, S.; Ilieva, K.

1997-01-01

The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

Effect of air condition on AP-1000 containment cooling performance in station black out accident

International Nuclear Information System (INIS)

Hendro Tjahjono

2015-01-01

AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transferred. The simulation results showed a decrease in power up to 5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10°C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. (author)

The Safety Assessment of OPR-1000 for Station Blackout Applying Combined Deterministic and Probabilistic Procedure

Energy Technology Data Exchange (ETDEWEB)

Kang, Dong Gu; Ahn, Seung-Hoon; Cho, Dae-Hyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

This is termed station blackout (SBO). However, it does not generally include the loss of available AC power to safety buses fed by station batteries through inverters or by alternate AC sources. Historically, risk analysis results have indicated that SBO was a significant contributor to overall core damage frequency. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident, which is a typical beyond design basis accident and important contributor to overall plant risk, is performed by applying the combined deterministic and probabilistic procedure (CDPP). In addition, discussions are made for reevaluation of SBO risk at OPR-1000 by eliminating excessive conservatism in existing PSA. The safety assessment of OPR-1000 for SBO accident, which is a typical BDBA and significant contributor to overall plant risk, was performed by applying the combined deterministic and probabilistic procedure. However, the reference analysis showed that the CDF and CCDP did not meet the acceptable risk, and it was confirmed that the SBO risk should be reevaluated. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it was demonstrated that the proposed CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

Genotype Imputation for Latinos Using the HapMap and 1000 Genomes Project Reference Panels

Directory of Open Access Journals (Sweden)

Xiaoyi eGao

2012-06-01

Full Text Available Genotype imputation is a vital tool in genome-wide association studies (GWAS and meta-analyses of multiple GWAS results. Imputation enables researchers to increase genomic coverage and to pool data generated using different genotyping platforms. HapMap samples are often employed as the reference panel. More recently, the 1000 Genomes Project resource is becoming the primary source for reference panels. Multiple GWAS and meta-analyses are targeting Latinos, the most populous and fastest growing minority group in the US. However, genotype imputation resources for Latinos are rather limited compared to individuals of European ancestry at present, largely because of the lack of good reference data. One choice of reference panel for Latinos is one derived from the population of Mexican individuals in Los Angeles contained in the HapMap Phase 3 project and the 1000 Genomes Project. However, a detailed evaluation of the quality of the imputed genotypes derived from the public reference panels has not yet been reported. Using simulation studies, the Illumina OmniExpress GWAS data from the Los Angles Latino Eye Study and the MACH software package, we evaluated the accuracy of genotype imputation in Latinos. Our results show that the 1000 Genomes Project AMR+CEU+YRI reference panel provides the highest imputation accuracy for Latinos, and that also including Asian samples in the panel can reduce imputation accuracy. We also provide the imputation accuracy for each autosomal chromosome using the 1000 Genomes Project panel for Latinos. Our results serve as a guide to future imputation-based analysis in Latinos.

Comparison of optical spectra recorded during DPF-1000U plasma experiments with gas-puffing

Directory of Open Access Journals (Sweden)

Zaloga Dobromil R.

2015-06-01

Full Text Available The results are presented of the optical spectra measurements for free plasma streams generated with the use of the modified DPF-1000U machine. This facility was recently equipped with a gas injection system (the so-called gas-puff placed on the symmetry axis behind the central opening in the inner electrode. The DPF-1000U experimental chamber was filled up with pure deuterium at the initial pressure of 1.6 or 2.4 mbar. Additionally, when the use was made of the gas-puff system about 1 cm3 of pure deuterium was injected at the pressure of 2 bars. The gas injection was initiated 1.5 or 2 ms before the triggering of the main discharge. The investigated plasma discharges were powered from a condenser bank charged initially to 23 kV (corresponding to the energy of 352 kJ, and the maximum discharge current amounted to about 1.8 MA. In order to investigate properties of a dense plasma column formed during DPF-1000U discharges the use was made of the optical emission spectroscopy. The optical spectra were recorded along the line of sight perpendicular to the vacuum chamber, using a Mechelle®900 spectrometer. The recent analysis of all the recorded spectra made it possible to compare the temporal changes in the electron density of a freely propagating plasma stream for discharges without and with the gas-puffing. Using this data an appropriate mode of operation of the DPF-1000U facility could be determined.

Power system EMC and IEC publication 1000

International Nuclear Information System (INIS)

Teichmann, H.

1997-01-01

A classification of the principal phenomena causing electromagnetic disturbances has been prepared by the International Electrotechnical Commission (IEC). Electromagnetic compatibility was defined (by the IEC) as 'the ability of a device, equipment or system to function satisfactorily in its electromagnetic environment without introducing intolerable electromagnetic disturbances in that environment'. The classification includes the following phenomena relative to power systems: (1) low-frequency phenomena, (2) radiated low-frequency phenomena, (2) conducted high-frequency phenomena, (3) radiated high-frequency phenomena, and (4) electrostatic discharge phenomena. The structure of the IEC publication 1000 was explained. Parts 1, 2, 3 and 6 of the publication were highlighted. The issues covered in those parts were: (Part 1) general, (Part 2) environment, (Part 3) limits, and (Part 6) generic standards. 6 refs

Recent ion measurements within the modified DPF-1000U facility

Directory of Open Access Journals (Sweden)

Kwiatkowski Roch

2015-06-01

Full Text Available In this note we describe measurements of ion beams emitted along the z-axis of the DPF-1000U facility operated at 23 kV, 334 kJ, and with the initial deuterium pressure of 1.6–2 hPa. The DPF-1000U device was recently renewed and equipped with a dynamic gas-puff valve placed inside the inner electrode. The investigated ions were recorded by means of ion pinhole cameras equipped with solid state nuclear track detectors of the PM-355® (PADC type. The energy spectra of ions were determined using a Thomson spectrometer placed on the symmetry axis at a distance of 160 cm from the electrodes outlets. The ion images recorded during discharges performed under different experimental conditions show that the ion beams have a complex structure, usually in the form of a central bunch and an annular stream composed of many micro-beams. Energies of the registered deuterons have been in the range of 30–700 keV, while the fast protons (which originated from the hydrogen remnants had energies in the range of 300–850 keV.

Correction of the Caulobacter crescentus NA1000 genome annotation.

Directory of Open Access Journals (Sweden)

Bert Ely

Full Text Available Bacterial genome annotations are accumulating rapidly in the GenBank database and the use of automated annotation technologies to create these annotations has become the norm. However, these automated methods commonly result in a small, but significant percentage of genome annotation errors. To improve accuracy and reliability, we analyzed the Caulobacter crescentus NA1000 genome utilizing computer programs Artemis and MICheck to manually examine the third codon position GC content, alignment to a third codon position GC frame plot peak, and matches in the GenBank database. We identified 11 new genes, modified the start site of 113 genes, and changed the reading frame of 38 genes that had been incorrectly annotated. Furthermore, our manual method of identifying protein-coding genes allowed us to remove 112 non-coding regions that had been designated as coding regions. The improved NA1000 genome annotation resulted in a reduction in the use of rare codons since noncoding regions with atypical codon usage were removed from the annotation and 49 new coding regions were added to the annotation. Thus, a more accurate codon usage table was generated as well. These results demonstrate that a comparison of the location of peaks third codon position GC content to the location of protein coding regions could be used to verify the annotation of any genome that has a GC content that is greater than 60%.

77 FR 74696 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on AP-1000...

Science.gov (United States)

2012-12-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on AP-1000; Notice of Meeting The ACRS Subcommittee on AP-1000 will hold a meeting on January 18, 2013, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting will be open to public...

SR-1000 radiofrequency chemo-hyperthermia for recurrent and metastatic peritoneo-pelvic malignant tumors

International Nuclear Information System (INIS)

Luo Jingwei; Xiong Jinghong; Xu Guozhen; Yu Zihao; Li Yexiong; Yin Weibo

2002-01-01

Objective: To evaluate the efficacy and tolerance of intraperitoneal chemo-hyperthermia (IPCH) with SR-1000 radiofrequency (RF) for recurrent or metastatic peritoneo-pelvic malignant tumors. Methods: Twenty-one patients with recurrent or metastatic peritoneo-pelvic malignant tumors received chemo-hyperthermia, with 9 having local pain and 14 having ascites. The Karnofsky scores were 40-80. After abdominal cavity aspiration and infusion of hot NS and chemotherapeutic agents, the temperature of abdominal cavity was increased and maintained at 40.5-42.5 degree C for 60-90 minutes with SR-1000 RF. Hyperthermia was given twice per week and chemotherapy once per week, with the whole treatment lasting for 2-4 weeks. The commonly used drugs were DDP, MMC, 5-FU and so on. Results: Local pain was relieved in 8 of 9 patients, complete disappearance of ascites in 10 of 14. The common side-effects were fat necrosis (14.3%) and abdominal pain (24.8%). Conclusions: Intraperitoneal chemo-hyperthermia with SR-1000 RF appears to be a promising new approach for patients with recurrent or metastatic peritoneo-pelvic malignant tumors, especially for those who did not response to systemic chemotherapy or whose tumor recurred after chemotherapy. As to bulky lesions, local supplementary radiotherapy should be given in order to obtain better local control

Strategy for 100-year life of the ACR-1000 concrete containment structure

International Nuclear Information System (INIS)

Abrishami, H.; Elgohary, M.

2006-01-01

The purpose of this paper is to present the Plant Life Management (PLiM) strategy for the concrete containment structure of the ACR-1000 (Advanced CANDU Reactor) designed by AECL. The ACR-1000 is designed for 100-year plant life including 60-year operating life and additional 40-year decommissioning period of time. The approach adopted for the PLiM strategy of the concrete containment structure is a preventive one, key areas being: 1) design methodology, 2) material performance and 3) life cycle management and ageing management program. In the design phase, in addition to strength and serviceability, durability is a major requirement during the service life and decommissioning phase of the ACR structure. Parameters affecting durability design include: a) concrete performance, b) structural application, and c) environmental conditions. Due to the complex nature of the environmental effects acting on structures during the service life of project, it is considered that true improved performance during the service life can be achieved by improving the material characteristics. Many recent innovations in advanced concrete materials technology have made it possible to produce modern concrete such as high-performance concrete with exceptional performance characteristics. In this paper, the PLiM strategy for the ACR-1000 concrete containment is presented. In addition to addressing the design methodology and material performance areas, a systematic approach for ageing management program for the concrete containment structure is presented. (author)

Design and performance evaluation of a 1000-year evapotranspiration-capillary surface barrier

Energy Technology Data Exchange (ETDEWEB)

Zhang, Zhuanfang Fred; Strickland, Christopher E.; Link, Steven O.

2017-02-01

Surface barrier technology is used to isolate radioactive waste and to reduce or eliminate recharge water to the waste zone for 1000 years or longer. However, the design and evaluation of such a barrier is challenging because of the extremely long design life. The Prototype Hanford Barrier (PHB) was designed as a 1000-year barrier with pre-determined design and performance objectives and demonstrated in field from 1994 to present. The barrier was tested to evaluate surface-barrier design and performance at the field scale under conditions of enhanced and natural precipitation and of no vegetation. The monitoring data demonstrate that the barrier satisfied nearly all key objectives. The PHB far exceeded the Resource Conservation and Recovery Act criteria, functioned in Hanford’s semiarid climate, limited drainage to well below the 0.5 mm yr-1 performance criterion, limited runoff, and minimized erosion. Given the two-decade record of successful performance and consideration of all the processes and mechanisms that could degrade the stability and hydrology in the future, the results suggest the PHB is very likely to perform for its 1000-year design life. This conclusion is based on two assumptions: (1) the exposed subgrade receives protection against erosion and (2) institutional controls prevent inadvertent human activity at the barrier. The PHB design can serve as the base for site-specific barriers over waste sites containing underground nuclear waste, uranium mine tailings, and hazardous mine waste.

AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-10-15

Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

Eakad peaksid saama 1000 krooni lisatoetust / Kadi Pärnits

Index Scriptorium Estoniae

Pärnits, Kadi, 1965-

2005-01-01

Ilmunud ka: Vooremaa 15. dets., lk. 2, Võrumaa Teataja 15. dets., lk. 2, Põhjarannik 15. dets., lk. 2, Severnoje Poberezhje 15. dets., lk. 2, Virumaa Teataja 16. dets., lk. 7, Vali Uudised 16. dets. lk. 2, Meie Maa 17. dets. lk. 2, Virumaa Nädalaleht 16. dets. lk. 3, Hiiu Leht 23. dets. lk. 2, Pärnu Postimees 14. jaan. 2006 lk. 19. Sotsiaaldemokraatliku Erakonna eelnõust hüvitada pensionäride retseptiravimiste ostmise kulu kuni 1000 krooni eest aastas

D-Wave's Approach to Quantum Computing: 1000-qubits and Counting!

CERN Multimedia

CERN. Geneva

2017-01-01

In this talk I will describe D-Wave's approach to quantum computing, including the system architecture of our 1000-qubit D-Wave 2X, its programming model, and performance benchmarks. Furthermore, I will describe how the native optimization and sampling capabilities of the quantum processor can be exploited to tackle problems in a variety of fields including medicine, machine learning, physics, and computational finance.

Economic Development Impact of 1,000 MW of Wind Energy in Texas

Energy Technology Data Exchange (ETDEWEB)

Reategui, S.; Hendrickson, S.

2011-08-01

Texas has approximately 9,727 MW of wind energy capacity installed, making it a global leader in installed wind energy. As a result of the significant investment the wind industry has brought to Texas, it is important to better understand the economic development impacts of wind energy in Texas. This report analyzes the jobs and economic impacts of 1,000 MW of wind power generation in the state. The impacts highlighted in this report can be used in policy and planning decisions and can be scaled to get a sense of the economic development opportunities associated with other wind scenarios. This report can also inform stakeholders in other states about the potential economic impacts associated with the development of 1,000 MW of new wind power generation and the relationships of different elements in the state economy.

Clinical and angiographic profile of patients with markedly elevated coronary calcium scores (≥1000) detected by electron beam computed tomography

International Nuclear Information System (INIS)

Almeda, Francis Q.; Shah, Rima; Senter, Shaun; Kason, Thomas T.; Haynie, Justin; Calvin, James E.; Kavinsky, Clifford J.; Snell, R. Jeffrey; Schaer, Gary L.; McLaughlin, Vallerie V.

2004-01-01

Objective: The objective of this study was to determine the clinical and angiographic profile of patients with extremely high coronary artery calcium scores (CACS; ≥1000) by electron beam computed tomography (EBCT). Methods: All patients at Rush University Medical Center who had a calcium score ≥1000 and a coronary angiogram performed from 1997 to 2002 were identified using a prospectively collected database. The baseline demographics, symptom status, and degree of coronary stenosis by angiography and subsequent rate of coronary intervention were compared with that of patients with calcium scores <1000. Results: The clinical and angiographic profile of patients with severe coronary calcification, detected by EBCT, revealed that patients with scores ≥1000 had a significantly higher prevalence of coronary stenosis ≥50% compared with patients with scores <1000 (97% vs. 57%, P<.001). The group with CACS ≥1000 was more likely to be male (90% vs. 75%, P=.027) and was older (64±8 vs. 59±10, P=.001) compared with the group with less severe calcification. Although there was a significantly higher rate of luminal stenosis detected by coronary angiography in the cohort with CACS ≥1000, there was no difference in subsequent percutaneous coronary intervention (PCI) and utilization of intracoronary stents between the two groups. Conclusions: A markedly elevated coronary calcium score (≥1000) is correlated with increasing age and is associated with an increased likelihood of coronary stenosis ≥50%. However, the decision to perform coronary angiography in patients with severe coronary calcification should not be based solely on these findings, but should remain primarily dependent on the degree of ischemia detected by clinical and functional assessment

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

2014-01-01

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

Evaluation of 500- and 1,000-mg doses of ciprofloxacin for the treatment of chancroid.

Science.gov (United States)

Bodhidatta, L; Taylor, D N; Chitwarakorn, A; Kuvanont, K; Echeverria, P

1988-01-01

A randomized, double-blind study was performed comparing ciprofloxacin in a 500-mg single dose with 1,000 mg (500-mg doses given 12 h apart) for the treatment of chancroid in Thailand. Haemophilus ducreyi was isolated from 87 (48%) of 180 men with a clinical diagnosis of chancroid. For men with ulcers that were culture positive for H. ducreyi, rates of cure were 100% in the 500-mg group and 98% in the 1,000-mg group. For men with ulcers that were culture negative for H. ducreyi, rates of cure were 93% in the 500-mg group and 96% in the 1,000-mg group. The MIC of ciprofloxacin for 50% of isolates among 85 isolates of H. ducreyi was 0.007 micrograms/ml (range, 0.002 to 0.03 micrograms/ml). No significant adverse effects were detected in either group. These data indicate that both of these treatment regimens are equally effective therapies for chancroid in Thailand. PMID:3293526

A procedure for temperature-stress fields calculation of WWER-1000 primary circuit in PTS event

Energy Technology Data Exchange (ETDEWEB)

Petkov, G [Technical Univ., Dept. Thermal and Nuclear Power Engineering, Sofia (Bulgaria); Groudev, P; Argirov, J [Bulgarian Academy of Science, Inst. for Nuclear Research and Nuclear Energy, Sofia (Bulgaria)

1997-09-01

The paper presents the procedure of an investigation of WWER-1000 primary circuit temperature-stress field by the use of thermohydraulic computation data for a pressurized thermal shock event ``Core overcooling``. The procedure is based on a model of the plane stress state with ideal contact between wall and medium for the calculation. The computation data are calculated on the base of WWER-1000 thermohydraulic model by the RELAP5/MOD3 codes. This model was developed jointly by the Bulgarian and BNL/USA staff to provide an analytical tool for performing safety analysis. As a result of calculations by codes the computation data for temperature field law (linear laws of a few distinguished parts) and pressure of coolant at points on inner surface of WWER-1000 primary circuit equipment are received. Such calculations can be used as a base for determination of all-important load-carrying sections of the primary circuit pipes and vessels, which need further consideration. (author). 7 refs, 2 figs, 2 tabs.

Minimum throttling feedwater control in VVER-1000 and PWR NPPs

International Nuclear Information System (INIS)

Symkin, B.E.; Thaulez, F.

2004-01-01

This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

1000 passwords exposed, what about yours?

CERN Multimedia

Computer Security Team

2011-01-01

In the last three issues of the Bulletin, we have stressed the importance of the secrecy of your password. Remember: Your password should be treated like a toothbrush: do not share it, and change it regularly! And this is not only valid for your CERN password, but also for any other password you use to log into your university or laboratory, Facebook or Twitter portals, or other web sites.   Unfortunately, recent security checks have revealed a huge area for improvement here. Within a period of only one week, more than 1000 different passwords passed through the CERN outer perimeter firewall in clear text. “Clear text” means that the password was readable to any adversary able to intercept the communication. Make sure your web connection is secure! This can easily be checked in the address bar of your web browser. If the address starts with “HTTPS”, everything is fine. If it is only “HTTP” (without “S&rd...

Confinement of antihydrogen for 1000 seconds

CERN Document Server

Andresen, G B; Baquero-Ruiz, M; Bertsche, W; Butler, E; Cesar, C L; Deller, A; Eriksson, S; Fajans, J; Friesen, T; Fujiwara, M C; Gill, D R; Gutierrez, A; Hangst, J S; Hardy, W N; Hayano, R S; Hayden, M E; Humphries, A J; Hydomako, R; Jonsell, S; Kemp, S; Kurchaninov, L; Madsen, N; Menary, S; Nolan, P; Olchanski, K; Olin, A; Pusa, P; Rasmussen, C Ø; Robicheaux, F; Sarid, E; Silveira, D M; So, C; Storey, J W; Thompson, R I; van der Werf, D P; Wurtele, J S; Yamazaki, Y

2011-01-01

Atoms made of a particle and an antiparticle are unstable, usually surviving less than a microsecond. Antihydrogen, made entirely of antiparticles, is believed to be stable, and it is this longevity that holds the promise of precision studies of matter-antimatter symmetry. We have recently demonstrated trapping of antihydrogen atoms by releasing them after a confinement time of 172 ms. A critical question for future studies is: how long can anti-atoms be trapped? Here we report the observation of anti-atom confinement for 1000 s, extending our earlier results by nearly four orders of magnitude. Our calculations indicate that most of the trapped anti-atoms reach the ground state. Further, we report the first measurement of the energy distribution of trapped antihydrogen which, coupled with detailed comparisons with simulations, provides a key tool for the systematic investigation of trapping dynamics. These advances open up a range of experimental possibilities, including precision studies of CPT symmetry and ...

Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

Energy Technology Data Exchange (ETDEWEB)

Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

2016-08-01

The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

25 CFR 1000.50 - What must a Tribe/Consortium seeking a planning grant submit in order to meet the planning phase...

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What must a Tribe/Consortium seeking a planning grant submit in order to meet the planning phase requirements? 1000.50 Section 1000.50 Indians OFFICE OF THE...) Planning and Negotiation Grants Advance Planning Grant Funding § 1000.50 What must a Tribe/Consortium...

Inclusion of Population-specific Reference Panel from India to the 1000 Genomes Phase 3 Panel Improves Imputation Accuracy.

Science.gov (United States)

Ahmad, Meraj; Sinha, Anubhav; Ghosh, Sreya; Kumar, Vikrant; Davila, Sonia; Yajnik, Chittaranjan S; Chandak, Giriraj R

2017-07-27

Imputation is a computational method based on the principle of haplotype sharing allowing enrichment of genome-wide association study datasets. It depends on the haplotype structure of the population and density of the genotype data. The 1000 Genomes Project led to the generation of imputation reference panels which have been used globally. However, recent studies have shown that population-specific panels provide better enrichment of genome-wide variants. We compared the imputation accuracy using 1000 Genomes phase 3 reference panel and a panel generated from genome-wide data on 407 individuals from Western India (WIP). The concordance of imputed variants was cross-checked with next-generation re-sequencing data on a subset of genomic regions. Further, using the genome-wide data from 1880 individuals, we demonstrate that WIP works better than the 1000 Genomes phase 3 panel and when merged with it, significantly improves the imputation accuracy throughout the minor allele frequency range. We also show that imputation using only South Asian component of the 1000 Genomes phase 3 panel works as good as the merged panel, making it computationally less intensive job. Thus, our study stresses that imputation accuracy using 1000 Genomes phase 3 panel can be further improved by including population-specific reference panels from South Asia.

24 CFR 1000.152 - How is the recipient to use criminal conviction information?

Science.gov (United States)

2010-04-01

... the criminal conviction information described in § 1000.150 only for applicant screening, lease... need for the information and who is an authorized officer, employee, or representative of the recipient...

SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

International Nuclear Information System (INIS)

Brettschuh, Werner

2006-01-01

The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

The phase system Fe-Ir-S at 1100, 1000 and 800 degree C

DEFF Research Database (Denmark)

Makovicky, Emil; Karup-Møller, Sven

1999-01-01

Phase relations in the dry condensed Fe-Ir-S system were determined at 1100, 1000 and 800 degrees C. Orientational runs were performed at 500 degrees C. Between 1100 and 800 degrees C, the system comprises five sulphides and an uninterrupted field of gamma(Fe, Ir). Fe1-xS dissolves 5.8 at.% Ir...... at 1100 degrees C, 3.4 at.% Ir at 1000 degrees C and 1.0 at.% Ir at 800 degrees C. The solubility of Fe in Ir2S3, IrS2 and IrSsimilar to 3 increases with decreasing temperature, reaching 2.5 at.% in the latter two sulphides at 800 degrees C. Thiospinel 'FeIr2S4' is nonstoichiometric, from Fe22.3Ir19.8S58...

Oxidation behavior of U-Si compounds in air from 25 to 1000 C

Science.gov (United States)

Sooby Wood, E.; White, J. T.; Nelson, A. T.

2017-02-01

The air oxidation behavior of U3Si2, USi, and U3Si5 is studied from room temperature to 1000 C. The onsets of breakaway oxidation for each compound are identified during synthetic air ramps to 1000 C using thermogravimetric analysis. Isothermal air oxidation tests are performed below and above the breakaway oxidation onset to discern the oxidation kinetic behavior of these candidate accident tolerant fuel forms. Uranium metal is tested in the same manner to provide a reference for the oxidation behavior. Thermogravimetric, x-ray diffraction, and scanning electron microscopy analysis are presented here along with a discussion of the oxidation behavior of these materials and the impact of the lack of oxidation resistance to their deployment as accident tolerant nuclear fuels.

24 CFR 1000.114 - How long does HUD have to review and act on a proposal to provide assistance to non low-income...

Science.gov (United States)

2010-04-01

... activity? 1000.114 Section 1000.114 Housing and Urban Development Regulations Relating to Housing and Urban... URBAN DEVELOPMENT NATIVE AMERICAN HOUSING ACTIVITIES Affordable Housing Activities § 1000.114 How long...

25 CFR 1000.246 - Must an AFA that contains a construction project or activity incorporate provisions of Federal...

Science.gov (United States)

2010-04-01

... as possible in the construction process. If Tribal construction standards are consistent with or... 25 Indians 2 2010-04-01 2010-04-01 false Must an AFA that contains a construction project or activity incorporate provisions of Federal construction standards? 1000.246 Section 1000.246 Indians OFFICE...

An evaluation of designed passive Core Makeup Tank (CMT) for China pressurized reactor (CPR1000)

International Nuclear Information System (INIS)

Wang, Mingjun; Tian, Wenxi; Qiu, Suizheng; Su, Guanghui; Zhang, Yapei

2013-01-01

Highlights: ► Only PRHRS is not sufficient to maintain reactor safety in case of SGTR accident. ► The Core Makeup Tank (CMT) is designed for CPR1000. ► Joint operation of PRHRS and CMT can keep reactor safety during the SGTR transient. ► CMT is a vital supplement for CPR1000 passive safety system design. - Abstract: Emergency Passive Safety System (EPSS) is an innovative design to improve reliability of nuclear power plants. In this work, the EPSS consists of secondary passive residual heat removal system (PRHRS) and the reactor Core Makeup Tank (CMT) system. The PRHRS, which has been studied in our previous paper, can effectively remove the core residual heat and passively improve the inherent safety by passive methods. The designed CMT, representing the safety improvement for CPR1000, is used to inject cool boron-containing water into the primary system during the loss of coolant accident. In this study, the behaviors of EPSS and transient characteristics of the primary loop system during the Steam Generator Tube Rupture (SGTR) accident are investigated using the nuclear reactor thermal hydraulic code RELAP5/MOD3.4. The results show that the designed CMT can protect the reactor primary loop from boiling and maintain primary loop coolant in single phase state. Both PRHRS and CMT operation ensures reactor safety during the SGTR accident. Results reported in this paper show that the designed CMT is a further safety improvement for CPR1000

Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

International Nuclear Information System (INIS)

Jordanov, M.

1999-01-01

The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

Lattice parameters and thermal expansion of delta-VNsub(1-x) from 298-1000 K

International Nuclear Information System (INIS)

Lengauer, W.; Ettmayer, P.

1986-01-01

The thermal expansion of VNsub(1-x) was determined from measurements of the lattice parameters in the temperature range of 298-1000 K and in the composition range of VNsub(0.707) - VNsub(0.996). Within the accuracy of the results the expansion of the lattice parameter with temperature is not dependent on the composition. The lattice parameter as a function of composition ([N]/[V] = 0.707-0.996) and temperature (198-1000 K) is given by a([N]/[V], T) = 0.38872+0.02488 ([N]/[V]) - (1.083+-0.021) x 10 -4 Tsup(1/2) + (6.2+-0.1) x 10 - sup6T. The coefficient of linear thermal expansion as a function of temperature (in the same range) is given by α(T) = a([N]/[V], T) -1 [(-5.04+-0.01) x 10 -5 Tsup(1/2) + (6.2+-0.1) x 10 -6 ]. The average linear thermal expansion coefficient is αsub(av) = 9.70 +- 0.15 x 10 -6 K -1 (298-1000 K). The data are compared with those of several fcc transition metal nitrides collected and evaluated from the literature. (Author)

Analysis of an accident with the main circulation tube rupture at the WWER-1000

International Nuclear Information System (INIS)

Boyadzhiev, A.I.; Stefanova, S.J.

1984-01-01

In connection with the forthcoming construction of a npp with the wwer-1000 reactor the loss of coolant accident associated with the main circulation tube rupture at the inlet near the reactor is analyzed. The relap4/mod6 program is used for the analysis. The data obtained show that the coolant outflow stage continues for about 25s. On the average the pressure in the circuits varies from 16 to 10 mpa per 0.1s and then it continues to decrease slowly. The pressure in the steam generator at the secondary circuits end increases approximately up to 6.9 MPa as a result of steam generator blocking and remaining coolant heating and then somewhat decreases owing to the primary circuit cooling. By the end of the fuel and can temperatures are equalized and the heat transfer coefficient is stabilized at the level of 100 w/1 (m 2 xK). It is concluded that during a loss of coolant accident at the wwer-1000 reactor in procesess of coolant blowdown in the medium power fuel elemets neither the fuel, melting temperature (3000 k), nor the critical temperature (1000 k) of plastic deformation zirconiu can initiation are attained

25 CFR 1000.20 - What is required in a planning report?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What is required in a planning report? 1000.20 Section....20 What is required in a planning report? As evidence that the Tribe/Consortium has completed the planning phase, the Tribe/Consortium must prepare and submit to the Secretary a final planning report. (a...

25 CFR 1000.70 - What criteria will the Director use to rank the applications and how many maximum points can be...

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What criteria will the Director use to rank the applications and how many maximum points can be awarded for each criterion? 1000.70 Section 1000.70 Indians... Process § 1000.70 What criteria will the Director use to rank the applications and how many maximum points...