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Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Virginia Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

2

Ohio Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Ohio nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

3

Arkansas Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

4

Michigan Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

5

Alabama Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

6

Texas Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

7

Tennessee Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Tennessee nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

8

Georgia Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

9

Nebraska Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Nebraska nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

10

Arizona Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

11

Maryland Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

12

Illinois Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Illinois nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

13

Florida Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Florida nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

14

Wisconsin Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Wisconsin nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

15

Minnesota Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Minnesota nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

16

CONSTRUCTION OF NUCLEAR POWER PLANTS  

E-Print Network [OSTI]

CONSTRUCTION OF NUCLEAR POWER PLANTS A Workshop on "NUCLEAR ENERGY RENAISSANCE" Addressing WAS DEEPLY INVOLVED IN ALMOST EVERY ASPECT OF BUILDING THE PLANTS THROUGH · Quality Assurance · Nuclear IN CONSTRUCTION OF ST. LUCIE-2 #12;LESSONS LEARNED FROM St. Lucie-2 NUCLEAR POWER PLANTS CAN BE BUILT

17

California Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

California nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

18

Pennsylvania Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Pennsylvania nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

19

Connecticut Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Connecticut nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

20

Nuclear Power Plant Design Project  

E-Print Network [OSTI]

Nuclear Power Plant Design Project A Response to the Environmental and Economic Challenge Of Global.............................................................................................................. 4 3. Assessment of the Issues and Needs for a New Plant

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

Louisiana Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Louisiana nuclear power plants, summer capacity and net generation, 2010" "Plant NameTotal Reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

22

Massachusetts Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

(percent)","Owner" "Pilgrim Nuclear Power Station Unit 1",685,"5,918",100.0,"Entergy Nuclear Generation Co" "1 Plant 1 Reactor",685,"5,918",100.0 "Note: Totals may not equal...

23

Owners of nuclear power plants  

SciTech Connect (OSTI)

Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of July 1996. The report is divided into sections representing different aspects of nuclear plant ownership.

Hudson, C.R.; White, V.S.

1996-11-01T23:59:59.000Z

24

North Carolina Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

Carolina nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

25

New Jersey Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

26

New York Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

27

South Carolina Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

South Carolina nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

28

Sabotage at Nuclear Power Plants  

SciTech Connect (OSTI)

Recently there has been a noted worldwide increase in violent actions including attempted sabotage at nuclear power plants. Several organizations, such as the International Atomic Energy Agency and the US Nuclear Regulatory Commission, have guidelines, recommendations, and formal threat- and risk-assessment processes for the protection of nuclear assets. Other examples are the former Defense Special Weapons Agency, which used a risk-assessment model to evaluate force-protection security requirements for terrorist incidents at DOD military bases. The US DOE uses a graded approach to protect its assets based on risk and vulnerability assessments. The Federal Aviation Administration and Federal Bureau of Investigation conduct joint threat and vulnerability assessments on high-risk US airports. Several private companies under contract to government agencies use formal risk-assessment models and methods to identify security requirements. The purpose of this paper is to survey these methods and present an overview of all potential types of sabotage at nuclear power plants. The paper discusses emerging threats and current methods of choice for sabotage--especially vehicle bombs and chemical attacks. Potential consequences of sabotage acts, including economic and political; not just those that may result in unacceptable radiological exposure to the public, are also discussed. Applicability of risk-assessment methods and mitigation techniques are also presented.

Purvis, James W.

1999-07-21T23:59:59.000Z

29

Nuclear power plants: structure and function  

SciTech Connect (OSTI)

Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety.

Hendrie, J.M.

1983-01-01T23:59:59.000Z

30

Organizational learning at nuclear power plants  

E-Print Network [OSTI]

The Nuclear Power Plant Advisory Panel on Organizational Learning provides channels of communications between the management and organization research projects of the MIT International Program for Enhanced Nuclear Power ...

Carroll, John S.

1991-01-01T23:59:59.000Z

31

Kansas Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

nuclear net generation (percent)","Owner" "Wolf Creek Generating Station Unit 1","1,160","9,556",100.0,"Wolf Creek Nuclear Optg Corp" "1 Plant 1 Reactor","1,160","9,556",100.0...

32

Vermont Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

mwh)","Share of State nuclear net generation (percent)","Owner" "Vermont Yankee Unit 1",620,"4,782",100.0,"Entergy Nuclear Vermont Yankee" "1 Plant 1 Reactor",620,"4,782",100.0...

33

Nuclear Power Plant Concrete Structures  

SciTech Connect (OSTI)

A nuclear power plant (NPP) involves complex engineering structures that are significant items of the structures, systems and components (SSC) important to the safe and reliable operation of the NPP. Concrete is the commonly used civil engineering construction material in the nuclear industry because of a number of advantageous properties. The NPP concrete structures underwent a great degree of evolution, since the commissioning of first NPP in early 1960. The increasing concern with time related to safety of the public and environment, and degradation of concrete structures due to ageing related phenomena are the driving forces for such evolution. The concrete technology underwent rapid development with the advent of chemical admixtures of plasticizer/super plasticizer category as well as viscosity modifiers and mineral admixtures like fly ash and silica fume. Application of high performance concrete (HPC) developed with chemical and mineral admixtures has been witnessed in the construction of NPP structures. Along with the beneficial effect, the use of admixtures in concrete has posed a number of challenges as well in design and construction. This along with the prospect of continuing operation beyond design life, especially after 60 years, the impact of extreme natural events ( as in the case of Fukushima NPP accident) and human induced events (e.g. commercial aircraft crash like the event of September 11th 2001) has led to further development in the area of NPP concrete structures. The present paper aims at providing an account of evolution of NPP concrete structures in last two decades by summarizing the development in the areas of concrete technology, design methodology and construction techniques, maintenance and ageing management of concrete structures.

Basu, Prabir [International Atomic Energy Agency (IAEA)] [International Atomic Energy Agency (IAEA); Labbe, Pierre [Electricity of France (EDF)] [Electricity of France (EDF); Naus, Dan [Oak Ridge National Laboratory (ORNL)] [Oak Ridge National Laboratory (ORNL)

2013-01-01T23:59:59.000Z

34

DOE Announces Loan Guarantee Applications for Nuclear Power Plant...  

Energy Savers [EERE]

Loan Guarantee Applications for Nuclear Power Plant Construction DOE Announces Loan Guarantee Applications for Nuclear Power Plant Construction October 2, 2008 - 3:43pm Addthis...

35

Maryland Nuclear Profile - Calvert Cliffs Nuclear Power Plant  

U.S. Energy Information Administration (EIA) Indexed Site

Calvert Cliffs Nuclear Power Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License...

36

New York Nuclear Profile - R E Ginna Nuclear Power Plant  

U.S. Energy Information Administration (EIA) Indexed Site

R E Ginna Nuclear Power Plant" "Unit","Summer Capacity (MW)","Net Generation (Thousand MWh)","Summer Capacity Factor (Percent)","Type","Commercial Operation Date","License...

37

Dose reduction at nuclear power plants  

SciTech Connect (OSTI)

The collective dose equivalent at nuclear power plants increased from 1250 rem in 1969 to nearly 54,000 rem in 1980. This rise is attributable primarily to an increase in nuclear generated power from 1289 MW-y to 29,155 MW-y; and secondly, to increased average plant age. However, considerable variation in exposure occurs from plant to plant depending on plant type, refueling, maintenance, etc. In order to understand the factors influencing these differences, an investigation was initiated to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at light water plants. Objectives are to: identify high-dose maintenance tasks and related dose-reduction techniques; investigate utilization of high-reliability, low-maintenance equipment; recommend improved radioactive waste handling equipment and procedures; examine incentives for dose reduction; and compile an ALARA handbook.

Baum, J.W.; Dionne, B.J.

1983-01-01T23:59:59.000Z

38

Floating nuclear power plant safety assurance principles  

SciTech Connect (OSTI)

In the north regions of the Russian federation and low density population areas, there is a real necessity for ecological clean energy small power sources. For this purpose, floating nuclear power plants, designed on the basis of atomic ship building engineering, are being conceptualized. It is possible to use the ship building plants for the reactor purposes. Issues such as radioactive waste management are described.

Zvonarev, B.M.; Kuchin, N.L.; Sergeev, I.V.

1993-12-31T23:59:59.000Z

39

Virtual environments for nuclear power plant design  

SciTech Connect (OSTI)

In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP).

Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W. [and others

1996-03-01T23:59:59.000Z

40

Washington Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Columbia Generating Station Unit...

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Iowa Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Duane Arnold Energy Center Unit...

42

SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS  

E-Print Network [OSTI]

SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS Piero Baraldi Chevalier EDF R&D ­ Simulation and information Technologies for Power generation system Department 6, Quai Monitoring, Empirical Modeling, Power Plants, Safety Critical Nuclear Instrumentation, Autoassociative models

Paris-Sud XI, Université de

43

Missouri Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Callaway Unit 1","1,190","8,996",100.0,"Union...

44

Mississippi Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Grand Gulf Unit 1","1,251","9,643",100.0,"Syste...

45

SELFMONITORING DISTRIBUTED MONITORING SYSTEM FOR NUCLEAR POWER PLANTS (PRELIMINARY VERSION)  

E-Print Network [OSTI]

SELF­MONITORING DISTRIBUTED MONITORING SYSTEM FOR NUCLEAR POWER PLANTS (PRELIMINARY VERSION) Aldo and identification are extremely important activities for the safety of a nuclear power plant. In particular inside huge and complex production plants. 1 INTRODUCTION Safety in nuclear power plants requires

46

Electromagnetic Compatibility in Nuclear Power Plants  

SciTech Connect (OSTI)

Electromagnetic compatibility (EMC) has long been a key element of qualification for mission critical instrumentation and control (I&C) systems used by the U.S. military. The potential for disruption of safety-related I&C systems by electromagnetic interference (EMI), radio-frequency interference (RFI), or power surges is also an issue of concern for the nuclear industry. Experimental investigations of the potential vulnerability of advanced safety systems to EMI/RFI, coupled with studies of reported events at nuclear power plants (NPPs) that are attributed to EMI/RFI, confirm the safety significance of EMC for both analog and digital technology. As a result, Oak Ridge National Laboratory has been engaged in the development of the technical basis for guidance that addresses EMC for safety-related I&C systems in NPPs. This research has involved the identification of engineering practices to minimize the potential impact of EMI/RFI and power surges and an evaluation of the ambient electromagnetic environment at NPPs to tailor those practices for use by the nuclear industry. Recommendations for EMC guidance have been derived from these research findings and are summarized in this paper.

Ewing, P.D.; Kercel, S.W.; Korsah, K.; Wood, R.T.

1999-08-29T23:59:59.000Z

47

Sensitivity analysis for the outages of nuclear power plants  

E-Print Network [OSTI]

Feb 17, 2012 ... Abstract: Nuclear power plants must be regularly shut down in order to perform refueling and maintenance operations. The scheduling of the ...

Kengy Barty

2012-02-17T23:59:59.000Z

48

Risk-informed incident management for nuclear power plants  

E-Print Network [OSTI]

Decision making as a part of nuclear power plant operations is a critical, but common, task. Plant management is forced to make decisions that may have safety and economic consequences. Formal decision theory offers the ...

Smith, Curtis Lee, 1966-

2002-01-01T23:59:59.000Z

49

Inspection of Nuclear Power Plant Containment Structures  

SciTech Connect (OSTI)

Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

Graves, H.L.; Naus, D.J.; Norris, W.E.

1998-12-01T23:59:59.000Z

50

Can New Nuclear Power Plants be Project Financed?  

E-Print Network [OSTI]

This paper considers the prospects for financing a wave of new nuclear power plants (NPP) using project financing, which is used widely in large capital intensive infrastructure investments, including the power and gas sectors, but has...

Taylor, Simon

51

Nuclear power plant performance assessment pertaining to plant aging in France and the United States  

E-Print Network [OSTI]

The effect of aging on nuclear power plant performance has come under increased scrutiny in recent years. The approaches used to make an assessment of this effect strongly influence the economics of nuclear power plant ...

Guyer, Brittany (Brittany Leigh)

2013-01-01T23:59:59.000Z

52

Some aspects of the decommissioning of nuclear power plants  

SciTech Connect (OSTI)

The major factors influencing the choice of a national concept for the decommissioning of nuclear power plants are examined. The operating lifetimes of power generating units with nuclear reactors of various types (VVER-1000, VVER-440, RBMK-1000, EGP-6, and BN-600) are analyzed. The basic approaches to decommissioning Russian nuclear power plants and the treatment of radioactive waste and spent nuclear fuel are discussed. Major aspects of the ecological and radiation safety of personnel, surrounding populations, and the environment during decommissioning of nuclear installations are identified.

Khvostova, M. S., E-mail: marinakhvostova@list.ru [St. Petersburg State Maritime Technical University (Sevmashvtuz), Severodvinsk Branch (Russian Federation)

2012-03-15T23:59:59.000Z

53

Seismic requirements for design of nuclear power plants and nuclear test facilities  

SciTech Connect (OSTI)

This standard establishes engineering requirements for the design of nuclear power plants and nuclear test facilities to accommodate vibratory effects of earthquakes.

Not Available

1985-02-01T23:59:59.000Z

54

UNSUPERVISED CLUSTERING FOR FAULT DIAGNOSIS IN NUCLEAR POWER PLANT COMPONENTS  

E-Print Network [OSTI]

1 UNSUPERVISED CLUSTERING FOR FAULT DIAGNOSIS IN NUCLEAR POWER PLANT COMPONENTS Piero Baraldi1 of prototypical behaviors. Its performance is tested with respect to an artificial case study and then applied on transients originated by different faults in the pressurizer of a nuclear power reactor. Key Words: Fault

Boyer, Edmond

55

Risk Framework for the Next Generation Nuclear Power Plant Construction  

E-Print Network [OSTI]

sector projects, and recently elevated to Best Practice status. However, its current format is inadequate to address the unique challenges of constructing the next generation of nuclear power plants (NPP). To understand and determine the risks...

Yeon, Jaeheum 1981-

2012-12-11T23:59:59.000Z

56

Mapping complexity sources in nuclear power plant domains  

E-Print Network [OSTI]

Understanding the sources of complexity in advanced Nuclear Power Plant (NPP) control rooms and their effects on human reliability is critical for ensuring safe performance of both operators and the entire system. New ...

Sasangohar, Farzan

57

Development of decontamination techniques for decommissioning commercial nuclear power plants  

SciTech Connect (OSTI)

NUPEC has been developing various techniques to safely and efficiently decommission large commercial nuclear power plants. The development work, referred to as the verification tests, has been performed since 1982. The verification tests on decontamination techniques have focused on the reduction of both occupational radiation exposure and radioactive waste volume. Experiments on various decontamination methods have been carried out. Prospects of applying efficient decontamination techniques to commercial nuclear power plant decommissioning are bright due to the experimental results.

Ishikura, T.; Miwa, T.; Onozawa, T.; Ohtsuka, H. [Nuclear Power Engineering Corp., Tokyo (Japan). Plant and Components Dept.; Ishigure, K. [Univ. of Tokyo (Japan). Dept. of Quantum Engineering and System Science

1993-12-31T23:59:59.000Z

58

Regulatory guidance for lightning protection in nuclear power plants  

SciTech Connect (OSTI)

Oak Ridge National Laboratory (ORNL) was engaged by the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) to develop the technical basis for regulatory guidance to address design and implementation practices for lightning protection systems in nuclear power plants (NPPs). Lightning protection is becoming increasingly important with the advent of digital and low-voltage analog systems in NPPs. These systems have the potential to be more vulnerable than older analog systems to the resulting power surges and electromagnetic interference (EMI) when lightning strikes facilities or power lines. This paper discusses the technical basis for guidance to licensees and applicants covered in Regulatory Guide (RG) 1.204, Guidelines for Lightning Protection of Nuclear Power Plants, issued August 2005. RG 1.204 describes guidance for practices that are acceptable to the NRC staff for protecting nuclear power structures and systems from direct lightning strikes and the resulting secondary effects. (authors)

Kisner, R. A.; Wilgen, J. B.; Ewing, P. D.; Korsah, K. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6007 (United States); Antonescu, C. E. [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

2006-07-01T23:59:59.000Z

59

Regulatory Guidance for Lightning Protection in Nuclear Power Plants  

SciTech Connect (OSTI)

Abstract - Oak Ridge National Laboratory (ORNL) was engaged by the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) to develop the technical basis for regulatory guidance to address design and implementation practices for lightning protection systems in nuclear power plants (NPPs). Lightning protection is becoming increasingly important with the advent of digital and low-voltage analog systems in NPPs. These systems have the potential to be more vulnerable than older analog systems to the resulting power surges and electromagnetic interference (EMI) when lightning strikes facilities or power lines. This paper discusses the technical basis for guidance to licensees and applicants covered in Regulatory Guide (RG) 1.204, Guidelines for Lightning Protection of Nuclear Power Plants, issued August 2005. RG 1.204 describes guidance for practices that are acceptable to the NRC staff for protecting nuclear power structures and systems from direct lightning strikes and the resulting secondary effects.

Kisner, Roger A [ORNL; Wilgen, John B [ORNL; Ewing, Paul D [ORNL; Korsah, Kofi [ORNL; Antonescu, Christina E [ORNL

2006-01-01T23:59:59.000Z

60

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSINGEmergency Planning for Nuclear Power Plants Determination ofproposed nuclear power plants . . . . . . . . . • . . . .

Yen, W.W.S.

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Aging of concrete structures in nuclear power plants  

SciTech Connect (OSTI)

The Structural Aging (SAG) Program, sponsored by the US Nuclear Regulatory Commission (USNRC) and conducted by the Oak Ridge National Laboratory (ORNL), had the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant structures for continued service. The program consists of three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued service determinations. Major accomplishments under the SAG Program during the first two years of its planned five-year duration have included: development of a Structural Materials Information Center and formulation of a Structural Aging Assessment Methodology for Concrete Structures in Nuclear Power Plants. 9 refs.

Naus, D.J.; Pland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))

1991-01-01T23:59:59.000Z

62

Aging management of containment structures in nuclear power plants  

SciTech Connect (OSTI)

Research is being conducted by ORNL under US Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques. assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.

Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [The Johns Hopkins Univ., Baltimore, MD (United States); Graves, H.L. III; Norris, W.E. [US Nuclear Regulatory Commission, Washington, DC (United States)

1994-12-31T23:59:59.000Z

63

Boron control system for a nuclear power plant  

SciTech Connect (OSTI)

Ion exchangers which reversibly store borate ions in a temperature dependent process are combined with evaporative boric acid recovery apparatus to provide a boron control system for controlling the reactivity of nuclear power plants. A plurality of ion exchangers are operated sequentially to provide varying amounts of boric acid to a nuclear reactor for load follow operations. Evaporative boric acid recovery apparatus is utilized for major changes in the boron concentration within the nuclear reactor.

Brown, W.W.; Van der Schoot, M.R.

1980-09-30T23:59:59.000Z

64

Ground-based testing of space nuclear power plants  

SciTech Connect (OSTI)

Small nuclear power plants for space applications are evaluated according to their testability in this two part report. The first part introduces the issues involved in testing these power plants. Some of the concerns include oxygen embrittlement of critical components, the test environment, the effects of a vacuum environment on materials, the practically of racing an activated test chamber, and possible testing alternative the SEHPTR, king develop at the Idaho National Engineering Laboratory. 10 refs., 6 figs., 1 tab.

McDonald, T.G.

1990-10-22T23:59:59.000Z

65

Nuclear Power Plant NDE Challenges - Past, Present, and Future  

SciTech Connect (OSTI)

The operating fleet of U.S. nuclear power plants was built to fossil plant standards (of workmanship, not fitness for service) and with good engineering judgment. Fortuitously, those nuclear power plants were designed using defense-in-depth concepts, with nondestructive examination (NDE) an important layer, so they can tolerate almost any component failure and still continue to operate safely. In the 30+ years of reactor operation, many material failures have occurred. Unfortunately, NDE has not provided the reliability to detect degradation prior to initial failure (breaching the pressure boundary). However, NDE programs have been improved by moving from prescriptive procedures to performance demonstrations that quantify inspection effectiveness for flaw detection probability and sizing accuracy. Other improvements include the use of risk-informed strategies to ensure that reactor components contributing the most risk receive the best and most frequent inspections. Another challenge is the recent surge of interest in building new nuclear power plants in the United States to meet increasing domestic energy demand. New construction will increase the demand for NDE but also offers the opportunity for more proactive inspections. This paper reviews the inception and evolution of NDE for nuclear power plants over the past 40 years, recounts lessons learned, and describes the needs remaining as existing plants continue operation and new construction is contemplated.

Doctor, S. R. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352 (United States)

2007-03-21T23:59:59.000Z

66

Nuclear power plant fire protection: philosophy and analysis. [PWR; BWR  

SciTech Connect (OSTI)

This report combines a fire severity analysis technique with a fault tree methodology for assessing the importance to nuclear power plant safety of certain combinations of components and systems. Characteristics unique to fire, such as propagation induced by the failure of barriers, have been incorporated into the methodology. By applying the resulting fire analysis technique to actual conditions found in a representative nuclear power plant, it is found that some safety and nonsafety areas are both highly vulnerable to fire spread and impotant to overall safety, while other areas prove to be of marginal importance. Suggestions are made for further experimental and analytical work to supplement the fire analysis method.

Berry, D. L.

1980-05-01T23:59:59.000Z

67

A methodology for evaluating ``new`` technologies in nuclear power plants  

SciTech Connect (OSTI)

As obsolescence and spare parts issues drive nuclear power plants to upgrade with new technology (such as optical fiber communication systems), the ability of the new technology to withstand stressors present where it is installed needs to be determined. In particular, new standards may be required to address qualification criteria and their application to the nuclear power plants of tomorrow. This paper discusses the failure modes and age-related degradation mechanisms of fiber optic communication systems, and suggests a methodology for identifying when accelerated aging should be performed during qualification testing.

Korsah, K.; Clark, R.L.; Holcomb, D.E.

1994-06-01T23:59:59.000Z

68

US nuclear power plant operating cost and experience summaries  

SciTech Connect (OSTI)

NUREG/CR-6577, U.S. Nuclear Power Plant Operating Cost and Experience Summaries, has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants. Cost incurred after initial construction are characterized as annual production costs, representing fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operating summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from annual operating reports submitted by the licensees, plant histories contained in Nuclear Power Experience, trade press articles, and the Nuclear Regulatory Commission (NRC) web site (www.nrc.gov).

Kohn, W.E.; Reid, R.L.; White, V.S.

1998-02-01T23:59:59.000Z

69

Report on aging of nuclear power plant reinforced concrete structures  

SciTech Connect (OSTI)

The Structural Aging Program provides the US Nuclear Regulatory Commission with potential structural safety issues and acceptance criteria for use in continued service assessments of nuclear power plant safety-related concrete structures. The program was organized under four task areas: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technology, and Quantitative Methodology for Continued Service Determinations. Under these tasks, over 90 papers and reports were prepared addressing pertinent aspects associated with aging management of nuclear power plant reinforced concrete structures. Contained in this report is a summary of program results in the form of information related to longevity of nuclear power plant reinforced concrete structures, a Structural Materials Information Center presenting data and information on the time variation of concrete materials under the influence of environmental stressors and aging factors, in-service inspection and condition assessments techniques, repair materials and methods, evaluation of nuclear power plant reinforced concrete structures, and a reliability-based methodology for current and future condition assessments. Recommendations for future activities are also provided. 308 refs., 61 figs., 50 tabs.

Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States); Ellingwood, B.R. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering

1996-03-01T23:59:59.000Z

70

New Hampshire Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA) Indexed Site

total reactors","Summer capacity (nw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Seabrook Unit 1","1,247","10,910",100.0,"NextEr...

71

Enhancement of NRC station blackout requirements for nuclear power plants  

SciTech Connect (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50, Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)

McConnell, M. W. [United States Nuclear Regulatory Commission, Mail Stop: 012-H2, Washington, DC 20555 (United States)

2012-07-01T23:59:59.000Z

72

Radioactive Effluents from Nuclear Power Plants Annual Report 2007  

SciTech Connect (OSTI)

This report describes radioactive effluents from commercial nuclear power plants (NPPs) in the United States. This information was reported by the licensees for radioactive discharges that occurred in 2007. The report provides information relevant to the potential impact of NPPs on the environment and on public health.

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation

2010-12-10T23:59:59.000Z

73

Radioactive Effluents from Nuclear Power Plants Annual Report 2008  

SciTech Connect (OSTI)

This report describes radioactive effluents from commercial nuclear power plants (NPPs) in the United States. This information was reported by the licensees for radioactive discharges that occurred in 2008. The report provides information relevant to the potential impact of NPPs on the environment and on public health.

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation

2010-12-10T23:59:59.000Z

74

Nuclear Power Plant Containment Pressure Boundary Research  

SciTech Connect (OSTI)

Research to address aging of the containment pressure boundary in light-water reactor plants is summarized. This research is aimed at understanding the significant factors relating occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containment and liners of concrete containment. This understanding will lead to improvements in risk-informed regulatory decision making. Containment pressure boundary components are described and potential aging factors identified. Quantitative tools for condition assessments of aging structures to maintain an acceptable level of reliability over the service life of the plant are discussed. Finally, the impact of aging (i.e., loss of shell thickness due to corrosion) on steel containment fragility for a pressurized water reactor ice-condenser plant is presented.

Cherry, J.L.; Chokshi, N.C.; Costello, J.F.; Ellingwood, B.R.; Naus, D.J.

1999-09-15T23:59:59.000Z

75

Mapping Complexity Sources in Nuclear Power Plant Domains Understanding the sources of complexity in advanced Nuclear Power Plant (NPP) control rooms and their  

E-Print Network [OSTI]

Mapping Complexity Sources in Nuclear Power Plant Domains Understanding the sources of complexity in advanced Nuclear Power Plant (NPP) control rooms and their effects on human reliability is critical of complexity leveraging network theory. INTRODUCTION The nuclear power industry in United States has declined

Cummings, Mary "Missy"

76

ASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM-OF-SYSTEMS  

E-Print Network [OSTI]

by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe stateASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant

Paris-Sud XI, Université de

77

Fiber optic sensors for nuclear power plant applications  

SciTech Connect (OSTI)

Studies have been carried out for application of Raman Distributed Temperature Sensor (RDTS) in Nuclear Power Plants (NPP). The high temperature monitoring in sodium circuits of Fast Breeder Reactor (FBR) is important. It is demonstrated that RDTS can be usefully employed in monitoring sodium circuits and in tracking the percolating sodium in the surrounding insulation in case of any leak. Aluminum Conductor Steel Reinforced (ACSR) cable is commonly used as overhead power transmission cable in power grid. The suitability of RDTS for detecting defects in ACSR overhead power cable, is also demonstrated.

Kasinathan, Murugesan; Sosamma, Samuel; BabuRao, Chelamchala; Murali, Nagarajan; Jayakumar, Tammana [Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu-603102 (India)

2012-05-17T23:59:59.000Z

78

Infrastructure development assistance modeling for nuclear power plant  

SciTech Connect (OSTI)

The purpose of this paper is to develop a model, a general frame to be utilized in assisting newcomer countries to start a nuclear power program. A nuclear power plant project involves technical complexity and high level of investment with long duration. Considering newcomers are mostly developing countries that lack the national infrastructure, key infrastructure issues may constitute the principal constraints to the development of a nuclear power program. In this regard, it is important to provide guidance and support to set up an appropriate infrastructure when we help them with the first launch of nuclear power plant project. To date, as a sole nuclear power generation company, KHNP has been invited many times to mentor or assist newcomer countries for their successful start of a nuclear power program since Republic of Korea is an exemplary case of a developing country which began nuclear power program from scratch and became a major world nuclear energy country in a short period of time. Through hosting events organized to aid newcomer countries' initiation of nuclear power projects, difficulties have been recognized. Each event had different contents according to circumstances because they were held as an unstructured and one-off thing. By developing a general model, we can give more adequate and effective aid in an efficient way. In this paper, we created a model to identify necessary infrastructures at the right stage, which was mainly based on a case of Korea. Taking into account the assistance we received from foreign companies and our own efforts for technological self-reliance, we have developed a general time table and specified activities required to do at each stage. From a donor's perspective, we explored various ways to help nuclear infrastructure development including technical support programs, training courses, and participating in IAEA technical cooperation programs on a regular basis. If we further develop the model, the next task would be to make the model more sophisticated as a 'semi-tailored model' so that it can be applied to a certain country reflecting its unique conditions. In accordance with its degree of established infrastructure, we can adjust or modify the model. Despite lots of benefits of using this model, there remain limitations such as time and budget constraints. These problems, however, can be addressed by cooperating with international organization such as the IAEA and other companies that share the same goal of helping newcomer countries introduce nuclear power. (authors)

Park, J. H.; Hwang, K.; Park, K. M.; Kim, S. W.; Lee, S. M. [Korea Hydro and Nuclear Power Co., LTD, 23, 106 gil, Yeongdong-daero, Gangnam-gu, 153-791 (Korea, Republic of)

2012-07-01T23:59:59.000Z

79

Tour of Entergy's Nuclear Power Plant in River Bend Owner: Entergy Gulf States Inc.  

E-Print Network [OSTI]

Tour of Entergy's Nuclear Power Plant in River Bend Owner: Entergy Gulf States Inc. Reactor Type a nuclear power plant. Plant was Entergy, a Boiling Water Reactor (BWR) type. Built in the 80's, it has of the veteran plant workers. The presentation gave the nuclear plant engineering basics and built

Ervin, Elizabeth K.

80

Aging management guideline for commercial nuclear power plants - heat exchangers  

SciTech Connect (OSTI)

This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in commercial nuclear power plant heat exchangers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Booker, S.; Lehnert, D.; Daavettila, N.; Palop, E.

1994-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

A Roadmap to Deploy New Nuclear Power Plants in the United States...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010: Volume II, Main Report A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010: Volume...

82

Understanding the nature of nuclear power plant risk  

SciTech Connect (OSTI)

This paper describes the evolution of understanding of severe accident consequences from the non-mechanistic assumptions of WASH-740 to WASH-1400, NUREG-1150, SOARCA and today in the interpretation of the consequences of the accident at Fukushima. As opposed to the general perception, the radiological human health consequences to members of the Japanese public from the Fukushima accident will be small despite meltdowns at three reactors and loss of containment integrity. In contrast, the radiation-related societal impacts present a substantial additional economic burden on top of the monumental task of economic recovery from the nonnuclear aspects of the earthquake and tsunami damage. The Fukushima accident provides additional evidence that we have mis-characterized the risk of nuclear power plant accidents to ourselves and to the public. The human health risks are extremely small even to people living next door to a nuclear power plant. The principal risk associated with a nuclear power plant accident involves societal impacts: relocation of people, loss of land use, loss of contaminated products, decontamination costs and the need for replacement power. Although two of the three probabilistic safety goals of the NRC address societal risk, the associated quantitative health objectives in reality only address individual human health risk. This paper describes the types of analysis that would address compliance with the societal goals. (authors)

Denning, R. S. [Ohio State Univ., 201 West 19th Avenue, Columbus, OH 43210-1142 (United States)

2012-07-01T23:59:59.000Z

83

U.S. Nuclear Power Plants: Continued Life or Replacement After 60? (released in AEO2010)  

Reports and Publications (EIA)

Nuclear power plants generate approximately 20% of U.S. electricity, and the plants in operation today are often seen as attractive assets in the current environment of uncertainty about future fossil fuel prices, high construction costs for new power plants (particularly nuclear plants), and the potential enactment of greenhouse gas regulations. Existing nuclear power plants have low fuel costs and relatively high power output. However, there is uncertainty about how long they will be allowed to continue operating.

2010-01-01T23:59:59.000Z

84

A Verification Framework for FBD based Software in Nuclear Power Plants Junbeom Yoo  

E-Print Network [OSTI]

A Verification Framework for FBD based Software in Nuclear Power Plants Junbeom Yoo Div, conducted using a nuclear power plant shutdown system being developed in Korea, demonstrated in nuclear power plant's reactor protection systems. The software verification framework uses two different

85

Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications  

E-Print Network [OSTI]

C4.2 Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications and control (I&C) systems play a crucial role in the operation of nuclear power plants (NPP) and other safety is available. The use of model checking to verify two nuclear power plant related systems is described: an arc

Heljanko, Keijo

86

Nuclear Power Plant Components Condition Monitoring by Probabilistic Support Vector , Redouane Seraouib  

E-Print Network [OSTI]

Nuclear Power Plant Components Condition Monitoring by Probabilistic Support Vector Machine Jie.zio@ecp.fr Abstract In this paper, an approach for the prediction of the condition of Nuclear Power Plant (NPP monitoring, Nuclear power plant, Point prediction hal-00790421,version1-12Jun2013 Author manuscript

Boyer, Edmond

87

PLC-Based Safety Critical Software Development for Nuclear Power Plants  

E-Print Network [OSTI]

PLC-Based Safety Critical Software Development for Nuclear Power Plants Junbeom Yoo1 , Sungdeok Cha development technique for nuclear power plants'I&C soft- ware controllers. To improve software safety, we in developing safety-critical control software for a Korean nuclear power plant, and experience to date has been

88

Condition monitoring of motor-operated valves in nuclear power plants Pierre Granjon  

E-Print Network [OSTI]

Condition monitoring of motor-operated valves in nuclear power plants Pierre Granjon Gipsa of nuclear power plants. Unfortunately, today's policies present a major drawback. Indeed, these monitoring safety constraints: nuclear power plants. Key components of such systems are motor-operated valves (MOVs

Boyer, Edmond

89

Vulnerability Analysis of a Nuclear Power Plant Considering Detonations of Explosive Devices  

E-Print Network [OSTI]

Vulnerability Analysis of a Nuclear Power Plant Considering Detonations of Explosive Devices Marko threats to a nuclear power plant in the year 1991 and after the 9/11 events in 2001. The methodology which strength and injuries of human beings with nuclear power plant models used in probabilistic safety

Cizelj, Leon

90

Childhood leukaemia incidence below the age of 5 years near French nuclear power plants  

E-Print Network [OSTI]

Childhood leukaemia incidence below the age of 5 years near French nuclear power plants D Laurier 1 living in the vicinity of nuclear power plants in Germany. We present herein results about the incidence of childhood leukaemia in the vicinity of nuclear power plants in France for the same age range. These results

Paris-Sud XI, Université de

91

Innovative applications of technology for nuclear power plant productivity improvements  

SciTech Connect (OSTI)

The nuclear power industry in several countries is concerned about the ability to maintain high plant performance levels due to aging and obsolescence, knowledge drain, fewer plant staff, and new requirements and commitments. Current plant operations are labor-intensive due to the vast number of operational and support activities required by commonly used technology in most plants. These concerns increase as plants extend their operating life. In addition, there is the goal to further improve performance while reducing human errors and increasingly focus on reducing operations and maintenance costs. New plants are expected to perform more productively than current plants. In order to achieve and increase high productivity, it is necessary to look at innovative applications of modern technologies and new concepts of operation. The Electric Power Research Inst. is exploring and demonstrating modern technologies that enable cost-effectively maintaining current performance levels and shifts to even higher performance levels, as well as provide tools for high performance in new plants. Several modern technologies being explored can provide multiple benefits for a wide range of applications. Examples of these technologies include simulation, visualization, automation, human cognitive engineering, and information and communications technologies. Some applications using modern technologies are described. (authors)

Naser, J. A. [Electric Power Research Inst., 3420 Hillview Avenue, Palo Alto, CA 94303 (United States)

2012-07-01T23:59:59.000Z

92

Decommissioning nuclear power plants - the wave of the future  

SciTech Connect (OSTI)

The paper discusses the project controls developed in the decommissioning of a nuclear power plant. Considerations are given to the contaminated piping and equipment that have to be removed and the spent and used fuel that has to be disposed of. The storage issue is of primary concern here. The cost control aspects and the dynamics of decommissioning are discussed. The effects of decommissioning laws on the construction and engineering firms are mentioned. 5 refs.

Griggs, F.S. Jr. [Raytheon Engineers and Contractors, Cumberland City, TN (United States)

1994-12-31T23:59:59.000Z

93

IEEE TRANSACTIONS ON NUCLEAR SCIENCE, VOL. 58, NO. 1, FEBRUARY 2011 277 Anomaly Detection in Nuclear Power Plants via  

E-Print Network [OSTI]

in Nuclear Power Plants via Symbolic Dynamic Filtering Xin Jin, Student Member, IEEE, Yin Guo, Soumik Sarkar detection algorithm for condition monitoring of nuclear power plants, where symbolic feature extraction Innova- tive & Secure (IRIS) simulator of nuclear power plants, and its per- formance is evaluated

Ray, Asok

94

Review of maintenance personnel practices at nuclear power plants  

SciTech Connect (OSTI)

As part of the Nuclear Regulatory Commission (NRC) sponsored Maintenance Qualifications and Staffing Project, the Pacific Northwest Laboratory (PNL) has conducted a preliminary assessment of nuclear power plant (NPP) maintenance practices. As requested by the NRC, the following areas within the maintenance function were examined: personnel qualifications, maintenance training, overtime, shiftwork and staffing levels. The purpose of the assessment was to identify the primary safety-related problems that required further analysis before specific recommendations can be made on the regulations affecting NPP maintenance operations.

Chockie, A.D.; Badalamente, R.V.; Hostick, C.J.; Vickroy, S.C.; Bryant, J.L.; Imhoff, C.H.

1984-05-01T23:59:59.000Z

95

Power Plant Power Plant  

E-Print Network [OSTI]

Basin Center for Geothermal Energy at University of Nevada, Reno (UNR) 2 Nevada Geodetic LaboratoryStillwater Power Plant Wabuska Power Plant Casa Diablo Power Plant Glass Mountain Geothermal Area Lassen Geothermal Area Coso Hot Springs Power Plants Lake City Geothermal Area Thermo Geothermal Area

Tingley, Joseph V.

96

Human Factors Considerations in New Nuclear Power Plants: Detailed Analysis.  

SciTech Connect (OSTI)

This Nuclear Regulatory Commission (NRC) sponsored study has identified human-performance issues in new and advanced nuclear power plants. To identify the issues, current industry developments and trends were evaluated in the areas of reactor technology, instrumentation and control technology, human-system integration technology, and human factors engineering (HFE) methods and tools. The issues were organized into seven high-level HFE topic areas: Role of Personnel and Automation, Staffing and Training, Normal Operations Management, Disturbance and Emergency Management, Maintenance and Change Management, Plant Design and Construction, and HFE Methods and Tools. The issues where then prioritized into four categories using a 'Phenomena Identification and Ranking Table' methodology based on evaluations provided by 14 independent subject matter experts. The subject matter experts were knowledgeable in a variety of disciplines. Vendors, utilities, research organizations and regulators all participated. Twenty issues were categorized into the top priority category. This Brookhaven National Laboratory (BNL) technical report provides the detailed methodology, issue analysis, and results. A summary of the results of this study can be found in NUREG/CR-6947. The research performed for this project has identified a large number of human-performance issues for new control stations and new nuclear power plant designs. The information gathered in this project can serve as input to the development of a long-term strategy and plan for addressing human performance in these areas through regulatory research. Addressing human-performance issues will provide the technical basis from which regulatory review guidance can be developed to meet these challenges. The availability of this review guidance will help set clear expectations for how the NRC staff will evaluate new designs, reduce regulatory uncertainty, and provide a well-defined path to new nuclear power plant licensing.

OHara,J.; Higgins, J.; Brown, W.; Fink, R.

2008-02-14T23:59:59.000Z

97

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

in U. S. Conunercial Nuclear Power Plants", Report WASH-Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"

Nero, A.V.

2010-01-01T23:59:59.000Z

98

Aging management guideline for commercial nuclear power plants-pumps  

SciTech Connect (OSTI)

This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D. [MDC-Ogden Environmental and Energy Services, Southfield, MI (United States)

1994-03-01T23:59:59.000Z

99

Compiling Utility Requirements For New Nuclear Power Plant Project  

SciTech Connect (OSTI)

Teollisuuden Voima Oy (TVO) submitted in November 2000 to the Finnish Government an application for a Decision-in-Principle concerning the construction of a new nuclear power plant in Finland. The actual investment decision can be made first after a positive decision has been made by the Government and the Parliament. Parallel to the licensing process, technical preparedness has been upheld so that the procurement process can be commenced without delay, when needed. This includes the definition of requirements for the plant and preliminary preparation of bid inquiry specifications. The core of the technical requirements corresponds to the specifications presented in the European Utility Requirement (EUR) document, compiled by major European electricity producers. Quite naturally, an amount of modifications to the EUR document are needed that take into account the country- and site-specific conditions as well as the experiences gained in the operation of the existing NPP units. Along with the EUR-related requirements concerning the nuclear island and power generation plant, requirements are specified for scope of supply as well as for a variety of issues related to project implementation. (author)

Patrakka, Eero [Teollisuuden Voima Oy, 27160 Olkiluoto (Finland)

2002-07-01T23:59:59.000Z

100

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

S. Commercial Nuclear Power Plants. WASH-1400. October 1975.Content of for Nuclear Power Plants. Regulatory Guide 1.101.PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSING PROCESS

Yen, W.W.S.

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Float level switch for a nuclear power plant containment vessel  

DOE Patents [OSTI]

This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

Powell, James G. (Clifton Park, NY)

1993-01-01T23:59:59.000Z

102

Float level switch for a nuclear power plant containment vessel  

DOE Patents [OSTI]

This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures.

Powell, J.G.

1993-11-16T23:59:59.000Z

103

Identification of good practices in the operation of nuclear power plants  

E-Print Network [OSTI]

This work developed an approach to diagnose problems and identify good practices in the operation of nuclear power plants using the system dynamics technique. The research began with construction of the ORSIM (Nuclear Power ...

Chen, Haibo, 1975-

2005-01-01T23:59:59.000Z

104

An examination of the pursuit of nuclear power plant construction projects in the United States  

E-Print Network [OSTI]

The recent serious reconsideration of nuclear power as a means for U.S. electric utilities to increase their generation capacity provokes many questions regarding the achievable success of future nuclear power plant ...

Guyer, Brittany (Brittany Leigh)

2011-01-01T23:59:59.000Z

105

Use of fuel cells for improving on-site emergency power availability and reliability ad nuclear power plants  

E-Print Network [OSTI]

To assure safe shutdown of a nuclear power plant, there must always be reliable means of decay heat removal provided, in last resort, by an Emergency Core Cooling System (ECCS). Currently the majority of nuclear power ...

Akkaynak, Derya

2005-01-01T23:59:59.000Z

106

Underwater nuclear power plants: improved safety, environmental compatibility and efficiency  

SciTech Connect (OSTI)

The further development of nuclear power engineering depends on the creation of a new generation of nuclear power plant (NPP) projects that have a high degree of safety. Decisions ensuring secure NPP exploitation must be based on the possibility of eliminating or localizing accidents. Using environmental properties to achieve secure NPP exploitation and accident elimination leads to suggest the construction of NPPs in water. An efficient way to provide energy to remote coastal areas is through use of floatable construction of prefabricated units. Floatable construction raises the quality of works, reduces expenditures on industrial facilities, and facilities building conditions in districts with extreme climatic conditions. A type of NPP that is situated on a shelf with the reactor compartment placed at the sea bottom is proposed. The underwater location of the reactor compartment on the fixed depth allows the natural water environment conditions of natural hydrostatic pressure, heat transfer and circulation to provide NPP safety. An example of new concept for power units with under-water localization of the reactor compartment is provided by the double-block NPP in a VVER reactor.

Galustov, K.Z.; Abadjyan, K.A.; Pavlov, A.B.

1991-01-01T23:59:59.000Z

107

Nuclear power plant simulation facility evaluation methodology: handbook. Volume 1  

SciTech Connect (OSTI)

This report is Volume 1 of a two-part document which describes a project conducted to develop a methodology to evaluate the acceptability of nuclear power plant (NPP) simulation facilities for use in the simulator-based portion of NRC's operator licensing examination. The proposed methodology is to be utilized during two phases of the simulation facility life-cycle, initial simulator acceptance and recurrent analysis. The first phase is aimed at ensuring that the simulator provides an accurate representation of the reference NPP. There are two components of initial simulator evaluation: fidelity assessment and a direct determination of the simulation facility's adequacy for operator testing. The second phase is aimed at ensuring that the simulation facility continues to accurately represent the reference plant throughout the life of the simulator. Recurrent evaluation is comprised of three components: monitoring reference plant changes, monitoring the simulator's hardware, and examining the data from actual plant transients as they occur. Volume 1 is a set of guidelines which details the steps involved in the two life-cycle phases, presents an overview of the methodology and data collection requirements, and addresses the formation of the evaluation team and the preparation of the evaluation plan. 29 figs.

Laughery, K.R. Jr.; Carter, R.J.; Haas, P.M.

1986-01-01T23:59:59.000Z

108

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Report LBL-5287. "Power Plant Reliability-Availability andConunercial Nuclear Power Plants", Report WASH-1400 (NUREG-Standards for Nuclear Power Plants," by A.V. Nero and Y.C.

Nero, A.V.

2010-01-01T23:59:59.000Z

109

Study of seismic design bases and site conditions for nuclear power plants  

SciTech Connect (OSTI)

This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches.

Not Available

1980-04-01T23:59:59.000Z

110

Management of aging of nuclear power plant containment structures  

SciTech Connect (OSTI)

Research addressing aging management of nuclear power plant concrete and steel containment structures is summarized. Accomplishments related to concrete containment structures include formation of a materials` property database; an aging assessment methodology to identify critical structures and degradation factors; guidelines and evaluation criteria for use in condition assessments; and a time-dependent reliability-based methodology for condition assessments and estimations of future performance. Under the steel containments and liners activity, a degradation assessment methodology has been developed, mathematical models that describe time-dependent changes in the containment due to aggressive environmental factors have been identified, and statistical data supporting the use of these models in time-dependent reliability analysis have been summarized.

Naus, D.; Oland, C.B. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.; Ellingwood, B. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering; Norris, W.E.; Graves, H.L. III [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1998-06-01T23:59:59.000Z

111

Prognostics and Life Beyond 60 for Nuclear Power Plants  

SciTech Connect (OSTI)

Safe, secure, reliable and sustainable energy supply is vital for advanced and industrialized life styles. To meet growing energy demand there is interest in longer term operation (LTO) for the existing nuclear power plant fleet and enhancing capabilities in new build. There is increasing use of condition based maintenance (CBM) for active components and periodic in service inspection (ISI) for passive systems: there is growing interest in deploying on-line monitoring. Opportunities exist to move beyond monitoring and diagnosis based on pattern recognition and anomaly detection to and prognostics with the ability to provide an estimate of remaining useful life (RUL). The adoption of digital I&C systems provides a framework within which added functionality including on-line monitoring can be deployed, and used to maintain and even potentially enhance safety, while at the same time improving planning and reducing both operations and maintenance costs.

Leonard J. Bond; Pradeep Ramuhalli; Magdy S. Tawfik; Nancy J. Lybeck

2011-06-01T23:59:59.000Z

112

Conceivable new recycling of nuclear waste by nuclear power companies in their plants  

E-Print Network [OSTI]

We outline the basic principles and the needed experiments for a conceivable new recycling of nuclear waste by the power plants themselves to avoid its transportation and storage to a (yet unknown) dumping area. Details are provided in an adjoining paper and in patents pending.

Ruggero Maria Santilli

1997-04-09T23:59:59.000Z

113

Example G Cost of construction of nuclear power plants Description of data  

E-Print Network [OSTI]

1 Example G Cost of construction of nuclear power plants Description of data Table G.1 gives reactor (LWR) power plants constructed in USA. It is required to predict the capital cost involved in the construction of further LWR power plants. The notation used in Table G.1 is explained in Table G.2. The final 6

Reid, Nancy

114

Example G Cost of construction of nuclear power plants Description of data  

E-Print Network [OSTI]

Example G Cost of construction of nuclear power plants Description of data Table G.1 gives data) power plants constructed in USA. It is required to predict the capital cost involved in the construction of further LWR power plants. The notation used in Table G.1 is explained in Table G.2. The final 6 lines

Reid, Nancy

115

Aging assessment of surge protective devices in nuclear power plants  

SciTech Connect (OSTI)

An assessment was performed to determine the effects of aging on the performance and availability of surge protective devices (SPDs), used in electrical power and control systems in nuclear power plants. Although SPDs have not been classified as safety-related, they are risk-important because they can minimize the initiating event frequencies associated with loss of offsite power and reactor trips. Conversely, their failure due to age might cause some of those initiating events, e.g., through short circuit failure modes, or by allowing deterioration of the safety-related component(s) they are protecting from overvoltages, perhaps preventing a reactor trip, from an open circuit failure mode. From the data evaluated during 1980--1994, it was found that failures of surge arresters and suppressers by short circuits were neither a significant risk nor safety concern, and there were no failures of surge suppressers preventing a reactor trip. Simulations, using the ElectroMagnetic Transients Program (EMTP) were performed to determine the adequacy of high voltage surge arresters.

Davis, J.F.; Subudhi, M. [Brookhaven National Lab., Upton, NY (United States)] [Brookhaven National Lab., Upton, NY (United States); Carroll, D.P. [Florida Univ., Gainesville, FL (United States)] [Florida Univ., Gainesville, FL (United States)

1996-01-01T23:59:59.000Z

116

Incremental costs and optimization of in-core fuel management of nuclear power plants  

E-Print Network [OSTI]

This thesis is concerned with development of methods for optimizing the energy production and refuelling decision for nuclear power plants in an electric utility system containing both nuclear and fossil-fuelled stations. ...

Watt, Hing Yan

1973-01-01T23:59:59.000Z

117

Aging of safety class 1E transformers in safety systems of nuclear power plants  

SciTech Connect (OSTI)

This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.

Roberts, E.W.; Edson, J.L.; Udy, A.C. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1996-02-01T23:59:59.000Z

118

License Stewardship Approach to Commercial Nuclear Power Plant Decommissioning  

SciTech Connect (OSTI)

The paper explores both the conceptual approach to decommissioning commercial nuclear facilities using a license stewardship approach as well as the first commercial application of this approach. The license stewardship approach involves a decommissioning company taking control of a site and the 10 CFR 50 License in order to complete the work utilizing the established trust fund. In conclusion: The license stewardship approach is a novel way to approach the decommissioning of a retired nuclear power plant that offers several key advantages to all parties. For the owner and regulators, it provides assurance that the station will be decommissioned in a safe, timely manner. Ratepayers are assured that the work will be completed for the price they already have paid, with the decommissioning contractor assuming the financial risk of decommissioning. The contractor gains control of the assets and liabilities, the license, and the decommissioning fund. This enables the decommissioning contractor to control their work and eliminates redundant layers of management, while bringing more focus on achieving the desired end state - a restored site. (authors)

Daly, P.T.; Hlopak, W.J. [Commercial Services Group, EnergySolutions 1009 Commerce Park, Oak Ridge, TN (United States)

2008-07-01T23:59:59.000Z

119

Contract Specifications For Olkiluoto 3 Nuclear Power Plant  

SciTech Connect (OSTI)

The Finnish Parliament ratified in May 2002 the application for a Decision-in- Principle (DIP) that was submitted by Teollisuuden Voima Oy (TVO) in November 2000 concerning the construction of a new nuclear power plant in Finland (FIN5). The bid inquiries for FIN5 were sent out by TVO in September 2002, requesting the bids by the end of March 2003. A contract with the plant supplier was signed in December 2003, implying the construction of a PWR of type EPR (European Pressurised Water Reactor) in Olkiluoto, called Olkiluoto 3 NPP. The preparation of Bid Inquiry Specifications (BIS) was initiated simultaneously with the filing of the application for DIP. The compilation of BIS was an evolutionary process, starting with the collection of relevant reference material, proceeding through the development of technical, administrative and commercial requirements, and ending with the consolidation of all documentation to a package containing the complete BIS. An intensive bid evaluation process started immediately after receiving the bids, accompanied by negotiations with the supplier candidates. The final Contract Specifications (CS) were constituted on the basis of the BIS supplemented with information contained in the bid and the outcome of the contract negotiations. (author)

Patrakka, Eero [Teollisuuden Voima Oy, 27160 Olkiluoto (Finland)

2004-07-01T23:59:59.000Z

120

Leasing of Nuclear Power Plants With Using Floating Technologies  

SciTech Connect (OSTI)

The proposal to organize and realize the international program on leasing of Nuclear Power Plant (NPP) reactor compartments is brought to the notice of potential partners. The proposal is oriented to the construction of new NPPs or to replacement of worked-out reactor units of the NPPs in operation on the sites situated near water area and to the use of afloat technologies for construction, mounting and transportation of reactor units as a Reactor Compartment Block Module (RCBM). According to the offered project the RCBM is fabricated in factory conditions at the largest Russian defense shipbuilding plant - State Unitary Enterprise 'Industrial Association SEVMASHPREDPRIYATIE' (SEVMASH) in the city of Severodvinsk of the Arkhangelsk region. After completion of assembling, testing and preliminary licensing the RCBM is given buoyancy by means of hermetic sealing and using pontoons and barges. The RCBM delivery to the NPP site situated near water area is performed by sea route. The RCBM is brought to the place of its installation with the use of appropriate hydraulic structures (canals, shipping locks), then is lowered on the basement constructed beforehand and incorporated into NPP scheme, of which the components are installed in advance. Floating means can be detached from the RCBM and used repeatedly for other RCBMs. Further procedure of NPP commissioning and its operation is carried out according to traditional method by power company in the framework of RCBM leasing with enlisting the services of firm-manufacturer's specialists either to provide reactor plant operation and concomitant processes or to perform author's supervision of operation. After completion of lifetime and reactor unloading the RCBM is dismantled with using the same afloat technology and taken away from NPP site to sea area entirely, together with its structures (reactor vessel, heat exchangers, pumps, pipelines and other equipment). Then RCBM is transported by shipping route to a firm-manufacturer, for subsequent reprocessing, utilization and storage. Nuclear fuel and radioactive wastes are removed from NPP site also. Use of leasing method removes legal problems connected with the transportation of radioactive materials through state borders as the RCBM remains a property of the state-producer at all stages of its life cycle. (authors)

Kuznetsov, Yu.N.; Gabaraev, B.A.; Reshetov, V.A.; Moskin, V.A. [Federal State Unitary Enterprise, N.A. Dollezhal' Scientific-Research and Design Institute of Power Engineering (Russian Federation)

2002-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents  

E-Print Network [OSTI]

A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents C. MUN , L Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents C. MUN a , L. CANTREL a , C Accidents Majeurs (DPAM), CEN Cadarache - France 1 b Commissariat à l'Energie Atomique (CEA), Direction de l'Energie

Boyer, Edmond

122

Feature Extraction for Data-Driven Fault Detection in Nuclear Power Plants Xin Jin, Robert M. Edwards and Asok Ray  

E-Print Network [OSTI]

Feature Extraction for Data-Driven Fault Detection in Nuclear Power Plants Xin Jin, Robert M monitoring of nuclear power plants (NPP) is one of the key issues addressed in nuclear energy safety research is performed during each nuclear power plant refueling outage, which may not be cost effective [1

Ray, Asok

123

A holistic investigation of complexity sources in nuclear power plant control rooms  

E-Print Network [OSTI]

The nuclear power community in the United States is moving to modernize aging power plant control rooms as well as develop control rooms for new reactors. New generation control rooms, along with modernized control rooms, ...

Sasangohar, Farzan

2011-01-01T23:59:59.000Z

124

The Decommissioning of the Trino Nuclear Power Plant  

SciTech Connect (OSTI)

Following a referendum in Italy in 1987, the four Nuclear Power Plants (NPPs) owned and operated by the state utility ENEL were closed. After closing the NPPs, ENEL selected a ''safestore'' decommissioning strategy; anticipating a safestore period of some 40-50 years. This approach was consistent with the funds collected during plant operation, and was reinforced by the lack of both a waste repository and a set of national free release limits for contaminated materials in Italy. During 1999, twin decisions were made to privatize ENEL and to transform the nuclear division into a separate subsidiary of the ENEL group. This group was renamed Sogin and during the following year, ownership of the company was transferred to the Italian Treasury. On formation, Sogin was asked by the Italian government to review the national decommissioning strategy. The objective of the review was to move from a safestore strategy to a prompt decommissioning strategy, with the target of releasing all of the nuclear sites by 2020. It was recognized that this target was conditional upon the availability of a national LLW repository together with interim stores for both spent fuel and HLW by 2009. The government also agreed that additional costs caused by the acceleration of the decommissioning program would be considered as stranded costs. These costs will be recovered by a levy on the kWh price of electricity, a process established and controlled by the Regulator of the Italian energy sector. Building on the successful collaboration to develop a prompt decommissioning strategy for the Latina Magnox reactor (1), BNFL and Sogin agreed to collaborate on an in depth study for the prompt decommissioning of the Sogin PWR at Trino. BNFL is currently decommissioning six NPPs and is at an advanced stage of planning for two further units, having completed a full and rigorous exercise to develop Baseline Decommissioning Plans (BDP's) for these stations. The BDP exercise utilizes the full range of BNFL decommissioning experience and knowledge to develop a strategy, methodology and cost for the decommissioning of NPPs. Over the past year, a prompt decommissioning strategy for Trino has been developed. The strategy has been based on the principles of minimizing waste products that require long term storage, maximizing 'free release' materials and utilizing existing and regulatory approved technologies.

Brusa, L.; DeSantis, R.; Nurden, P. L.; Walkden, P.; Watson, B.

2002-02-27T23:59:59.000Z

125

Feasibility Study of Hydrogen Production at Existing Nuclear Power Plants  

SciTech Connect (OSTI)

Cooperative Agreement DE-FC07-06ID14788 was executed between the U.S. Department of Energy, Electric Transportation Applications, and Idaho National Laboratory to investigate the economics of producing hydrogen by electrolysis using electricity generated by nuclear power. The work under this agreement is divided into the following four tasks: Task 1 – Produce Data and Analyses Task 2 – Economic Analysis of Large-Scale Alkaline Electrolysis Task 3 – Commercial-Scale Hydrogen Production Task 4 – Disseminate Data and Analyses. Reports exist on the prospect that utility companies may benefit from having the option to produce electricity or produce hydrogen, depending on market conditions for both. This study advances that discussion in the affirmative by providing data and suggesting further areas of study. While some reports have identified issues related to licensing hydrogen plants with nuclear plants, this study provides more specifics and could be a resource guide for further study and clarifications. At the same time, this report identifies other area of risks and uncertainties associated with hydrogen production on this scale. Suggestions for further study in some of these topics, including water availability, are included in the report. The goals and objectives of the original project description have been met. Lack of industry design for proton exchange membrane electrolysis hydrogen production facilities of this magnitude was a roadblock for a significant period. However, recent design breakthroughs have made costing this facility much more accurate. In fact, the new design information on proton exchange membrane electrolyzers scaled to the 1 kg of hydrogen per second electrolyzer reduced the model costs from $500 to $100 million. Task 1 was delayed when the original electrolyzer failed at the end of its economic life. However, additional valuable information was obtained when the new electrolyzer was installed. Products developed during this study include a process model and a N2H2 economic assessment model (both developed by the Idaho National Laboratory). Both models are described in this report. The N2H2 model closely tracked and provided similar results as the H2A model and was instrumental in assessing the effects of plant availability on price when operated in the shoulder mode for electrical pricing. Differences between the H2A and N2H2 model are included in this report.

Stephen Schey

2009-07-01T23:59:59.000Z

126

Ice Thermal Storage Systems for Nuclear Power Plant Supplemental Cooling and Peak Power Shifting  

SciTech Connect (OSTI)

Availability of cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. One potential solution is to use ice thermal storage (ITS) systems that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses the ice for supplemental cooling during peak demand time. ITS also provides a way to shift a large amount of electricity from off peak time to peak time. For once-through cooling plants near a limited water body, adding ITS can bring significant economic benefits and avoid forced derating and shutdown during extremely hot weather. For the new plants using dry cooling towers, adding the ITS systems can effectively reduce the efficiency loss during hot weather so that new plants could be considered in regions lack of cooling water. This paper will review light water reactor cooling issues and present the feasibility study results.

Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

2013-03-01T23:59:59.000Z

127

Department of Mechanical and Nuclear Engineering Spring 2012 East Campus Power Plant Deaerator Optimization  

E-Print Network [OSTI]

PENNSTATE Department of Mechanical and Nuclear Engineering Spring 2012 East Campus Power Plant Deaerator Optimization Overview In the East Campus Power plant a new Deaerator system has been installed which Deaerator is the most efficient and then make a recommendation to the plant of which one should

Demirel, Melik C.

128

Confirmation of the seismic resistance of nuclear power plant equipment after assembly  

SciTech Connect (OSTI)

It is shown that the natural frequencies and damping decrements of nuclear power plant equipment can only be determined experimentally and directly at the power generation units (reactors) of nuclear power plants under real disassembly conditions for the equipment, piping network, thermal insulation, etc. A computational experimental method is described in which the natural frequencies and damping decrements are determined in the field and the seismic resistance is reevaluated using these values. This method is the basis of the standards document 'Methods for confirming the dynamic characteristics of systems and components of the generating units of nuclear power plants which are important for safety' prepared and introduced in 2012.

Kaznovsky, P. S.; Kaznovsky, A. P.; Saakov, E. S.; Ryasnyj, S. I. [JSC 'Atomtehenergo' (Russian Federation)

2013-05-15T23:59:59.000Z

129

Waste Minimization Policy at the Romanian Nuclear Power Plant  

SciTech Connect (OSTI)

The radioactive waste management system at Cernavoda Nuclear Power Plant (NPP) in Romania was designed to maintain acceptable levels of safety for workers and to protect human health and the environment from exposure to unacceptable levels of radiation. In accordance with terminology of the International Atomic Energy Agency (IAEA), this system consists of the ''pretreatment'' of solid and organic liquid radioactive waste, which may include part or all of the following activities: collection, handling, volume reduction (by an in-drum compactor, if appropriate), and storage. Gaseous and aqueous liquid wastes are managed according to the ''dilute and discharge'' strategy. Taking into account the fact that treatment/conditioning and disposal technologies are still not established, waste minimization at the source is a priority environmental management objective, while waste minimization at the disposal stage is presently just a theoretical requirement for future adopted technologies . The necessary operational and maintenance procedures are in place at Cernavoda to minimize the production and contamination of waste. Administrative and technical measures are established to minimize waste volumes. Thus, an annual environmental target of a maximum 30 m3 of radioactive waste volume arising from operation and maintenance has been established. Within the first five years of operations at Cernavoda NPP, this target has been met. The successful implementation of the waste minimization policy has been accompanied by a cost reduction while the occupational doses for plant workers have been maintained at as low as reasonably practicable levels. This paper will describe key features of the waste management system along with the actual experience that has been realized with respect to minimizing the waste volumes at the Cernavoda NPP.

Andrei, V.; Daian, I.

2002-02-26T23:59:59.000Z

130

Aging assessment of large electric motors in nuclear power plants  

SciTech Connect (OSTI)

Large electric motors serve as the prime movers to drive high capacity pumps, fans, compressors, and generators in a variety of nuclear plant systems. This study examined the stressors that cause degradation and aging in large electric motors operating in various plant locations and environments. The operating history of these machines in nuclear plant service was studied by review and analysis of failure reports in the NPRDS and LER databases. This was supplemented by a review of motor designs, and their nuclear and balance of plant applications, in order to characterize the failure mechanisms that cause degradation, aging, and failure in large electric motors. A generic failure modes and effects analysis for large squirrel cage induction motors was performed to identify the degradation and aging mechanisms affecting various components of these large motors, the failure modes that result, and their effects upon the function of the motor. The effects of large motor failures upon the systems in which they are operating, and on the plant as a whole, were analyzed from failure reports in the databases. The effectiveness of the industry`s large motor maintenance programs was assessed based upon the failure reports in the databases and reviews of plant maintenance procedures and programs.

Villaran, M.; Subudhi, M. [Brookhaven National Lab., Upton, NY (United States)

1996-03-01T23:59:59.000Z

131

Aging of Class 1E batteries in safety systems of nuclear power plants  

SciTech Connect (OSTI)

This report presents the results of a study of aging effects on safety-related batteries in nuclear power plants. The purpose is to evaluate the aging effects caused by operation within a nuclear facility and to evaluate maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant-Aging Research approach and investigates the materials used in battery construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes battery-failure events reported in various data bases, and evaluates recommended maintenance practices. Data bases that were analyzed included the NRC's Licensee Event Report system, the Institute for Nuclear Power Operations' Nuclear Plant Reliability Data System, the Oak Ridge National Laboratory's In-Plant Reliability Data System, and The S.M. Stoller Corporation's Nuclear Power Experience data base.

Edson, J.L.; Hardin, J.E.

1987-07-01T23:59:59.000Z

132

The impact of offsite factors on the safety performance of small nuclear power plants  

SciTech Connect (OSTI)

The results of an analysis of the influence of offsite factors on small nuclear power-plant (SNPP) safety performance during postulated severe accidents are presented. Given the plant locations in the immediate vicinity of residential areas and the impossibility of accomplishing the expeditious evacuation of the public, the risk caused by an SNPP severe accident may be considerably less than that for such an event in a large nuclear power plant. 3 refs., 3 figs., 5 tabs.

Baranaev, Yu.D.; Viktorov, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation)

1991-01-01T23:59:59.000Z

133

Nuclear Safety Risk Management in Refueling Outage of Qinshan Nuclear Power Plant  

SciTech Connect (OSTI)

The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclear safety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control. (authors)

Meijing Wu; Guozhang Shen [Qinshan Nuclear power company (China)

2006-07-01T23:59:59.000Z

134

Two novel procedures for aggregating randomized model ensemble outcomes for robust signal reconstruction in nuclear power plants monitoring systems  

E-Print Network [OSTI]

reconstruction in nuclear power plants monitoring systems P. Baraldi1 , E. Zio1,* , G. Gola2 , D. Roverso2 , M importance for the safe and reliable operation of nuclear power plants. Auto-associative regression models of nuclear power plants for it allows the timely detection of malfunctions and anomalies during operation

Paris-Sud XI, Université de

135

Abstract--Resins are used in nuclear power plants for water ultrapurification. Two approaches are considered in this work  

E-Print Network [OSTI]

Abstract--Resins are used in nuclear power plants for water ultrapurification. Two approaches in manufacturing ultrapure water for nuclear power plants. Resins allow the removal of ionic impurities to subparts-per-million. Thereby in nuclear power plants, resins contribute to guarantee personnel safety, to control feed system

Paris-Sud XI, Université de

136

DATA-DRIVEN ON-LINE PREDICTION OF THE AVAILABLE RECOVERY TIME IN NUCLEAR POWER PLANT FAILURE SCENARIOS  

E-Print Network [OSTI]

1 DATA-DRIVEN ON-LINE PREDICTION OF THE AVAILABLE RECOVERY TIME IN NUCLEAR POWER PLANT FAILURE-XADS). Key Words: Recovery Time, Emergency Accident Management, Nuclear Power Plant, Lead- Bismuth Eutectic e [Øwre, 2001]. Yet, the problem of what kind of decision support to provide to nuclear power plant

Boyer, Edmond

137

Conference committees Chairman of the conference "New Nuclear Power Plants in the Netherlands", June 21-22, 2011,  

E-Print Network [OSTI]

Conference committees Chairman of the conference "New Nuclear Power Plants in the Netherlands Nuclear Power Plants, September 15-19, 2003, Kyoto, Japan. Session chairman GENES4/ANP2003 ,,International Conference on Global Environment and Advanced Nuclear Power Plants, September 15-19, 2003, Kyoto

138

NUCLEAR PLANT AND CONTROL  

E-Print Network [OSTI]

NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: software require- ments, safety analysis, formal for the digital protection systems of a nuclear power plant. When spec- ifying requirements for software and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital

139

Atmospheric dispersion and the radiological consequences of normal airborne effluents from a nuclear power plant  

SciTech Connect (OSTI)

The relationship between the consequences of the normal exhaust of radioactive materials in air from nuclear power plants and atmospheric dispersion is studied. Because the source terms of the exhaust from a nuclear power plant are relatively low and their radiological consequences are far less than the corresponding authoritative limits, the atmospheric dispersion models, their various modifications, and selections of relevant parameters have few effects on those consequences. In the environmental assessment and siting, the emphasis should not be placed on the consequence evaluation of routine exhaust of nuclear power plants, and the calculation of consequences of the exhaust and atmospheric field measurements should be appropriately, simplified. 12 refs., 5 figs., 7 tabs.

Fang, D.; Yang, L. [Tsinghua Univ., Beijing (China); Sun, C.Z. [Suhou Nuclear Research Inst., Suzhou (China)

1995-01-01T23:59:59.000Z

140

Integration of Biorefineries and Nuclear Cogeneration Power Plants - A Preliminary Analysis  

SciTech Connect (OSTI)

Biomass-based ethanol and nuclear power are two viable elements in the path to U.S. energy independence. Numerous studies suggest nuclear power could provide a practical carbon-free heat source alternative for the production of biomass-based ethanol. In order for this coupling to occur, it is necessary to examine the interfacial requirements of both nuclear power plants and bioethanol refineries. This report describes the proposed characteristics of a small cogeneration nuclear power plant, a biochemical process-based cellulosic bioethanol refinery, and a thermochemical process-based cellulosic biorefinery. Systemic and interfacial issues relating to the co-location of either type of bioethanol facility with a nuclear power plant are presented and discussed. Results indicate future co-location efforts will require a new optimized energy strategy focused on overcoming the interfacial challenges identified in the report.

Greene, Sherrell R [ORNL; Flanagan, George F [ORNL; Borole, Abhijeet P [ORNL

2009-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Quiz # 7, STAT 383, Prof. Suman Sanyal, April 8, 2009 (Q2, Page 354) To decide whether the pipe welds in a nuclear power plant meet  

E-Print Network [OSTI]

welds in a nuclear power plant meet specifications, a random sample of welds is to be selected : µ nuclear power plants is to determine if welds

Sanyal, Suman

142

Probabilistic methods in seismic risk assessment for nuclear power plants: proceedings  

SciTech Connect (OSTI)

The state-of-the-art in seismic risk analysis applied to the design and siting of nuclear power plants was addressed in this meeting. Presentations were entered individually into the date base. (ACR)

Not Available

1983-01-01T23:59:59.000Z

143

Maximizing nuclear power plant performance via mega-uprates and subsequent license renewal  

E-Print Network [OSTI]

The goal of this thesis is to develop a methodology to evaluate the engineering and economic implications of maximizing performance of the United States' commercial fleet of nuclear power plants. This methodology addresses ...

DeWitte, Jacob D. (Jacob Dominic)

2014-01-01T23:59:59.000Z

144

Dynamic reliability using entry-time approach for maintenance of nuclear power plants  

E-Print Network [OSTI]

-time processes have the potential to provide a significantly greater range of applicability and flexibility than traditional reliability tools for case studies related to equipment and components in nuclear power plants. In this dissertation, the finite...

Wang, Shuwen

2009-05-15T23:59:59.000Z

145

Pacific Basin Nuclear Conference (PBNC 2012), BEXCO, Busan, Korea, March 18 ~ 23, 2012 CHALLENGES OF CYBER SECURITY FOR NUCLEAR POWER PLANTS  

E-Print Network [OSTI]

PBNC 2012 CHALLENGES OF CYBER SECURITY FOR NUCLEAR POWER PLANTS Kwangjo Kim KAIST, Daejeon, Korea.kim@kustar.ac.ae Abstract Nuclear Power Plants (NPPs) become one of the most important infrastructures in providing improvement. 1. Introduction Nuclear Power Plants (NPPs) become one of the most important infrastructures

Kim, Kwangjo

146

Prognostics Health Management and Life Beyond 60 for Nuclear Power Plants  

SciTech Connect (OSTI)

There is growing interest in longer-term operation of the current US nuclear power plant fleet. This paper will present an overview of prognostic health management (PHM) technologies that could play a role in the safe and effective operation of nuclear power plants during extended life. A case study in prognostics for materials degradation assessment, using laboratory-scale measurements, is briefly discussed, and technical gaps that need to be addressed prior to PHM system deployment for nuclear power life extension are presented.

Ramuhalli, Pradeep; Coble, Jamie B.; Meyer, Ryan M.; Bond, Leonard J.

2013-12-01T23:59:59.000Z

147

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"Densities Surrounding Nuclear Power Plants," by A.V. Nero,

Nero, jA.V.

2010-01-01T23:59:59.000Z

148

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Standards for Nuclear Power Plants," by A.V. Nero and Y.C.Planning for Nuclear Power Plants in California," by W.W.S.Surrounding Nuclear Power Plants," by A.V. Nero, C.H.

Nero, jA.V.

2010-01-01T23:59:59.000Z

149

Incentive regulation of investor-owned nuclear power plants by public utility regulators. Revision 1  

SciTech Connect (OSTI)

The US Nuclear Regulatory Commission (NRC) periodically surveys the Federal Energy Regulatory Commission (FERC) and state regulatory commissions that regulate utility owners of nuclear power plants. The NRC is interested in identifying states that have established economic or performance incentive programs applicable to nuclear power plants, how the programs are being implemented, and in determining the financial impact of the programs on the utilities. The NRC interest stems from the fact that such programs have the potential to adversely affect the safety of nuclear power plants. The current report is an update of NUREG/CR-5975, Incentive Regulation of Investor-Owned Nuclear Power Plants by Public Utility Regulators, published in January 1993. The information in this report was obtained from interviews conducted with each state regulatory agency that administers an incentive program and each utility that owns at least 10% of an affected nuclear power plant. The agreements, orders, and settlements that form the basis for each incentive program were reviewed as required. The interviews and supporting documentation form the basis for the individual state reports describing the structure and financial impact of each incentive program.

McKinney, M.D.; Seely, H.E.; Merritt, C.R.; Baker, D.C. [Pacific Northwest Lab., Richland, WA (United States)

1995-04-01T23:59:59.000Z

150

Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants - Final Technical Report  

SciTech Connect (OSTI)

OAK B188 Summary of methods proposed for risk informing the design and regulation of future nuclear power plants. All elements of the historical design and regulation process are preserved, but the methods proposed for new plants use probabilistic risk assessment methods as the primary decision making tool.

Ritterbusch, Stanley; Golay, Michael; Duran, Felicia; Galyean, William; Gupta, Abhinav; Dimitrijevic, Vesna; Malsch, Marty

2003-01-29T23:59:59.000Z

151

Guidance for Deployment of Mobile Technologies for Nuclear Power Plant Field Workers  

SciTech Connect (OSTI)

This report is a guidance document prepared for the benefit of commercial nuclear power plants’ (NPPs) supporting organizations and personnel who are considering or undertaking deployment of mobile technology for the purpose of improving human performance and plant status control (PSC) for field workers in an NPP setting. This document especially is directed at NPP business managers, Electric Power Research Institute, Institute of Nuclear Power Operations, and other non-Information Technology personnel. This information is not intended to replace basic project management practices or reiterate these processes, but is to support decision-making, planning, and preparation of a business case.

Heather D. Medema; Ronald K. Farris

2012-09-01T23:59:59.000Z

152

Power to the People or Regulatory Ratcheting? Explaining the Success (or Failure) of Attempts to Site Commercial U.S. Nuclear Power Plants: 1954 -19961  

E-Print Network [OSTI]

to Site Commercial U.S. Nuclear Power Plants: 1954 - 19961 7 April 2014 Eric Berndt2 and Daniel P. Aldrich to attempt siting nuclear power plant facilities in large numbers in the 1960s. By the late 1990s, more than 1984). In the case of the Shoreham Nuclear Generating Station in Long Island, the plant was completed

153

EDF Nuclear Power Plants Operating Experience with MOX fuel  

SciTech Connect (OSTI)

EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many details and finally no important impact is anticipated. The industrial maturity of plutonium recycling activities is fully demonstrated and a new progress can be done with a complete confidence. The licensing process of 'MOX Parity' core management is in progress and its implementation on the 20 PWR is now expected at mid 2007. (author)

Thibault, Xavier [EDF Generation, Tour EDF Part Dieu - 9 rue des Cuirassiers B.P.3181 - 69402 Lyon Cedex 03 (France)

2006-07-01T23:59:59.000Z

154

Inspection of Nuclear Power Plant Structures - Overview of Methods and Related Applications  

SciTech Connect (OSTI)

The objectives of this limited study were to provide an overview of the methods that are available for inspection of nuclear power plant reinforced concrete and metallic structures, and to provide an assessment of the status of methods that address inspection of thick, heavily-reinforced concrete and inaccessible areas of the containment metallic pressure boundary. In meeting these objectives a general description of nuclear power plant safety-related structures was provided as well as identification of potential degradation factors, testing and inspection requirements, and operating experience; methods for inspection of nuclear power plant reinforced concrete structures and containment metallic pressure boundaries were identified and described; and applications of nondestructive evaluation methods specifically related to inspection of thick-section reinforced concrete structures and inaccessible portions of containment metallic pressure boundaries were summarized. Recommendations are provided on utilization of test article(s) to further advance nondestructive evaluation methods related to thick-section, heavily-reinforced concrete and inaccessible portions of the metallic pressure boundary representative of nuclear power plant containments. Conduct of a workshop to provide an update on applications and needed developments for nondestructive evaluation of nuclear power plant structures would also be of benefit.

Naus, Dan J [ORNL

2009-05-01T23:59:59.000Z

155

Online Condition Monitoring to Enable Extended Operation of Nuclear Power Plants  

SciTech Connect (OSTI)

Safe, secure, and economic operation of nuclear power plants will remain of strategic significance. New and improved monitoring will likely have increased significance in the post-Fukushima world. Prior to Fukushima, many activities were already underway globally to facilitate operation of nuclear power plants beyond their initial licensing periods. Decisions to shut down a nuclear power plant are mostly driven by economic considerations. Online condition monitoring is a means to improve both the safety and economics of extending the operating lifetimes of nuclear power plants, enabling adoption of proactive aging management. With regard to active components (e.g., pumps, valves, motors, etc.), significant experience in other industries has been leveraged to build the science base to support adoption for online condition-based maintenance and proactive aging management in the nuclear industry. Many of the research needs are associated with enabling proactive management of aging in passive components (e.g., pipes, vessels, cables, containment structures, etc.). This paper provides an overview of online condition monitoring for the nuclear power industry with an emphasis on passive components. Following the overview, several technology/knowledge gaps are identified, which require addressing to facilitate widespread online condition monitoring of passive components.

Meyer, Ryan M.; Bond, Leonard J.; Ramuhalli, Pradeep

2012-03-31T23:59:59.000Z

156

Improved assessment of population doses and risk factors for a nuclear power plant under accident conditions  

E-Print Network [OSTI]

In order to assess the doses received by the members of the public due to an accident at a nuclear power plant, a number of physical processes must be modeled. These processes include the release of radioactive materials, the atmospheric dispersion... representative of the industry. Generic reactor sites must be conceptualized in order to obtain meteorologic data which is representative of the areas within the United States in which nuclear power facilities have been sited, Information such as population...

Meyer, Christopher Martin

1985-01-01T23:59:59.000Z

157

Initiating Event Rates at U.S. Nuclear Power Plants 1988–2013  

SciTech Connect (OSTI)

Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant’s low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC’s Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

John A. Schroeder; Gordon R. Bower

2014-02-01T23:59:59.000Z

158

Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report  

SciTech Connect (OSTI)

OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

NONE

2000-08-01T23:59:59.000Z

159

Reassessment of selected factors affecting siting of Nuclear Power Plants  

SciTech Connect (OSTI)

Brookhaven National Laboratory has performed a series of probabilistic consequence assessment calculations for nuclear reactor siting. This study takes into account recent insights into severe accident source terms and examines consequences in a risk based format consistent with the quantitative health objectives (QHOs) of the NRC`s Safety Goal Policy. Simplified severe accident source terms developed in this study are based on the risk insights of NUREG-1150. The results of the study indicate that both the quantity of radioactivity released in a severe accident as well as the likelihood of a release are lower than those predicted in earlier studies. The accident risks using the simplified source terms are examined at a series of generic plant sites, that vary in population distribution, meteorological conditions, and exclusion area boundary distances. Sensitivity calculations are performed to evaluate the effects of emergency protective action assumptions on the risk of prompt fatality and latent cancers fatality, and population relocation. The study finds that based on the new source terms the prompt and latent fatality risks at all generic sites meet the QHOs of the NRC`s Safety Goal Policy by margins ranging from one to more than three orders of magnitude. 4 refs., 17 figs., 24 tabs.

Davis, R.E.; Hanson, A.L.; Mubayi, V.; Nourbakhsh, H.P.

1997-02-01T23:59:59.000Z

160

OVERVIEW OF A RECONFIGURABLE SIMULATOR FOR MAIN CONTROL ROOM UPGRADES IN NUCLEAR POWER PLANTS  

SciTech Connect (OSTI)

This paper provides background on a reconfigurable control room simulator for nuclear power plants. The main control rooms in current nuclear power plants feature analog technology that is growing obsolete. The need to upgrade control rooms serves the practical need of maintainability as well as the opportunity to implement newer digital technologies with added functionality. There currently exists no dedicated research simulator for use in human factors design and evaluation activities for nuclear power plant modernization in the U.S. The new research simulator discussed in this paper provides a test bed in which operator performance on new control room concepts can be benchmarked against existing control rooms and in which new technologies can be validated for safety and usability prior to deployment.

Ronald L. Boring

2012-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

The status of nuclear power plants in the People's Republic of China  

SciTech Connect (OSTI)

China's main energy source is coal, but transportation and environmental problems make that fuel less than desirable. Therefore, the Chinese, as part of an effort toward alternative energy sources, are developing nuclear power plants. In addition to providing a cleaner power source, development of nuclear energy would improve the Chinese economic condition and give the nation greater world status. China's first plants, at Qinshan and Daya Bay, are still incomplete. However, China is working toward completion of those reactors and planning the training and operating procedures needed to operate them. At the same time, it is improving its nuclear fuel exports. As they develop the capability for generating nuclear power, the Chinese seem to be aware of the accompanying quality and safety considerations, which they have declared to be first priorities. 50 refs., 7 figs.

Puckett, J.

1991-05-01T23:59:59.000Z

162

Diversity Strategies for Nuclear Power Plant Instrumentation and Control Systems  

SciTech Connect (OSTI)

This report presents the technical basis for establishing acceptable mitigating strategies that resolve diversity and defense-in-depth (D3) assessment findings and conform to U.S. Nuclear Regulatory Commission (NRC) requirements. The research approach employed to establish appropriate diversity strategies involves investigation of available documentation on D3 methods and experience from nuclear power and nonnuclear industries, capture of expert knowledge and lessons learned, determination of best practices, and assessment of the nature of common-cause failures (CCFs) and compensating diversity attributes. The research described in this report does not provide guidance on how to determine the need for diversity in a safety system to mitigate the consequences of potential CCFs. Rather, the scope of this report provides guidance to the staff and nuclear industry after a licensee or applicant has performed a D3 assessment per NUREG/CR-6303 and determined that diversity in a safety system is needed for mitigating the consequences of potential CCFs identified in the evaluation of the safety system design features. Succinctly, the purpose of the research described in this report was to answer the question, 'If diversity is required in a safety system to mitigate the consequences of potential CCFs, how much diversity is enough?' The principal results of this research effort have identified and developed diversity strategies, which consist of combinations of diversity attributes and their associated criteria. Technology, which corresponds to design diversity, is chosen as the principal system characteristic by which diversity criteria are grouped to form strategies. The rationale for this classification framework involves consideration of the profound impact that technology-focused design diversity provides. Consequently, the diversity usage classification scheme involves three families of strategies: (1) different technologies, (2) different approaches within the same technology, and (3) different architectures within the same technology. Using this convention, the first diversity usage family, designated Strategy A, is characterized by fundamentally diverse technologies. Strategy A at the system or platform level is illustrated by the example of analog and digital implementations. The second diversity usage family, designated Strategy B, is achieved through the use of distinctly different technologies. Strategy B can be described in terms of different digital technologies, such as the distinct approaches represented by general-purpose microprocessors and field-programmable gate arrays. The third diversity usage family, designated Strategy C, involves the use of variations within a technology. An example of Strategy C involves different digital architectures within the same technology, such as that provided by different microprocessors (e.g., Pentium and Power PC). The grouping of diversity criteria combinations according to Strategies A, B, and C establishes baseline diversity usage and facilitates a systematic organization of strategic approaches for coping with CCF vulnerabilities. Effectively, these baseline sets of diversity criteria constitute appropriate CCF mitigating strategies for digital safety systems. The strategies represent guidance on acceptable diversity usage and can be applied directly to ensure that CCF vulnerabilities identified through a D3 assessment have been adequately resolved. Additionally, a framework has been generated for capturing practices regarding diversity usage and a tool has been developed for the systematic assessment of the comparative effect of proposed diversity strategies (see Appendix A).

Wood, Richard Thomas [ORNL; Belles, Randy [ORNL; Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Korsah, Kofi [ORNL; Loebl, Andy [ORNL; Mays, Gary T [ORNL; Muhlheim, Michael David [ORNL; Mullens, James Allen [ORNL; Poore III, Willis P [ORNL; Qualls, A L [ORNL; Wilson, Thomas L [ORNL; Waterman, Michael E. [U.S. Nuclear Regulatory Commission

2010-02-01T23:59:59.000Z

163

NUCLEAR PLANT OPERATIONS AND  

E-Print Network [OSTI]

NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: moderator temper ature coefficient, reactivity co reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed absorption cross-section behavior. Consequently, if NUCLEAR TECHNOLOGY VOL. 140 NOV. 2002 147 #12;Demazière

Demazière, Christophe

164

NUCLEAR PLANT OPERATIONS AND  

E-Print Network [OSTI]

NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: moderator temper- ature coefficient, reactivity co reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed. Consequently, if*E-mail: demaz@nephy.chalmers.se NUCLEAR TECHNOLOGY VOL. 140 NOV. 2002 147 #12;high-burnup fuel

Pázsit, Imre

165

The Regulatory Challenges of Decommissioning Nuclear Power Plants in Korea - 13101  

SciTech Connect (OSTI)

As of 2012, 23 units of nuclear power plants are in operation, but there is no experience of permanent shutdown and decommissioning of nuclear power plant in Korea. It is realized that, since late 1990's, improvement of the regulatory framework for decommissioning has been emphasized constantly from the point of view of International Atomic Energy Agency (IAEA)'s safety standards. And it is known that now IAEA prepare the safety requirement on decommissioning of facilities, its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to the result of IAEA's Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, it was recommended that the regulatory framework for decommissioning should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became also important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we identify the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA's safety standards in order to achieve our goal. And then the plan is to be established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. After dealing with it, it is expected that the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards. (authors)

Lee, Jungjoon; Ahn, Sangmyeon; Choi, Kyungwoo [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)] [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Juyoul; Kim, Juyub [FNC Technology, 46 Tabsil-ro, Giheung-gu, Yongin 446-902 (Korea, Republic of)] [FNC Technology, 46 Tabsil-ro, Giheung-gu, Yongin 446-902 (Korea, Republic of)

2013-07-01T23:59:59.000Z

166

Natural Circulation in Water Cooled Nuclear Power Plants Phenomena, models, and methodology for system reliability assessments  

SciTech Connect (OSTI)

In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''.

Jose Reyes

2005-02-14T23:59:59.000Z

167

Sensitivity analysis for the outages of nuclear power plants  

E-Print Network [OSTI]

Feb 17, 2012 ... Energy generation in France is a competitive market, whereas ... from wind farms, solar energy or run of river plant without pondage.

2012-02-17T23:59:59.000Z

168

Volume I, Summary Report: A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010:  

Broader source: Energy.gov [DOE]

Nuclear power plants in the United States currently produce about 20 percent of the nation’s electricity. This nuclear-generated electricity is safe, clean and economical, and does not emit...

169

State of the art review of radioactive waste volume reduction techniques for commercial nuclear power plants  

SciTech Connect (OSTI)

A review is made of the state of the art of volume reduction techniques for low level liquid and solid radioactive wastes produced as a result of: (1) operation of commercial nuclear power plants, (2) storage of spent fuel in away-from-reactor facilities, and (3) decontamination/decommissioning of commercial nuclear power plants. The types of wastes and their chemical, physical, and radiological characteristics are identified. Methods used by industry for processing radioactive wastes are reviewed and compared to the new techniques for processing and reducing the volume of radioactive wastes. A detailed system description and report on operating experiences follow for each of the new volume reduction techniques. In addition, descriptions of volume reduction methods presently under development are provided. The Appendix records data collected during site surveys of vendor facilities and operating power plants. A Bibliography is provided for each of the various volume reduction techniques discussed in the report.

Not Available

1980-04-01T23:59:59.000Z

170

Nuclear Power Plant NDE Challenges - Past, Present, and Future  

SciTech Connect (OSTI)

This is a paper that covers the major thrust of NDE work that PNNL has conducted for the U.S. Nuclear Regulatory Commission from 1977 to the present.

Doctor, Steven R.

2007-01-01T23:59:59.000Z

171

Identification of performance indicators for nuclear power plants  

E-Print Network [OSTI]

Performance indicators have been assuming an increasingly important role in the nuclear industry. An integrated methodology is proposed in this research for the identification and validation of performance indicators for ...

Sui, Yu, 1973-

2001-01-01T23:59:59.000Z

172

An analysis of nuclear power plant operating costs: A 1995 update  

SciTech Connect (OSTI)

Over the years real (inflation-adjusted) O&M cost have begun to level off. The objective of this report is to determine whether the industry and NRC initiatives to control costs have resulted in this moderation in the growth of O&M costs. Because the industry agrees that the control of O&M costs is crucial to the viability of the technology, an examination of the factors causing the moderation in costs is important. A related issue deals with projecting nuclear operating costs into the future. Because of the escalation in nuclear operating costs (and the fall in fossil fuel prices) many State and Federal regulatory commissions are examining the economics of the continued operation of nuclear power plants under their jurisdiction. The economics of the continued operation of a nuclear power plant is typically examined by comparing the cost of the plants continued operation with the cost of obtaining the power from other sources. This assessment requires plant-specific projections of nuclear operating costs. Analysts preparing these projections look at past industry-wide cost trends and consider whether these trends are likely to continue. To determine whether these changes in trends will continue into the future, information about the causal factors influencing costs and the future trends in these factors are needed. An analysis of the factors explaining the moderation in cost growth will also yield important insights into the question of whether these trends will continue.

NONE

1995-04-21T23:59:59.000Z

173

Community emergency response to nuclear power plant accidents: A selected and partially annotated bibliography  

SciTech Connect (OSTI)

The role of responding to emergencies at nuclear power plants is often considered the responsibility of the personnel onsite. This is true for most, if not all, of the incidents that may happen during the course of the plant`s operating lifetime. There is however, the possibility of a major accident occurring at anytime. Major nuclear accidents at Chernobyl and Three Mile Island have taught their respective countries and communities a significant lesson in local emergency preparedness and response. Through these accidents, the rest of the world can also learn a great deal about planning, preparing and responding to the emergencies unique to nuclear power. This bibliography contains books, journal articles, conference papers and government reports on emergency response to nuclear power plant accidents. It does not contain citations for ``onsite`` response or planning, nor does it cover the areas of radiation releases from transportation accidents. The compiler has attempted to bring together a sampling of the world`s collective written experience on dealing with nuclear reactor accidents on the sate, local and community levels. Since the accidents at Three Mile Island and Chernobyl, that written experience has grown enormously.

Youngen, G.

1988-10-01T23:59:59.000Z

174

Aerial Radiation Measurements from the Fukushima Dai-ichi Nuclear Power Plant Accident  

SciTech Connect (OSTI)

This document is a slide show type presentation concerning DOE and Aerial Measuring System (AMS) activities and results with respect to assessing the consequences of the releases from the Fukushima Dai-ichi Nuclear Power Plant. These include ground monitoring and aerial monitoring.

Guss, P. P.

2012-07-16T23:59:59.000Z

175

Survey of thermal-hydraulic models of commercial nuclear power plants  

SciTech Connect (OSTI)

A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC's current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described.

Determan, J.C.; Hendrix, C.E.

1992-12-01T23:59:59.000Z

176

Survey of thermal-hydraulic models of commercial nuclear power plants  

SciTech Connect (OSTI)

A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC`s current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described.

Determan, J.C.; Hendrix, C.E.

1992-12-01T23:59:59.000Z

177

Ranking of four potential nuclear power plant sites in Iraq according to the collective dose criterion  

SciTech Connect (OSTI)

The collective dose criterion was used to rank four potential nuclear power-plant sites. Baiji, Al-Mahzam, Al-Abbasia, and Abu-Dalaf. Atmospheric as well as aquatic releases of radionuclides into the environment from the VVER 440 nuclear power plant during normal operation were used to estimate the collective dose equivalents. The results indicated that the collective doses at Baiji, Al-Mahzam, Al-Abbasia, and Abu-Dalaf were 3.6 x 10{sup -2}, 4.7 x 10{sup -2}, 1.1 x 10{sup -1}, and 1.2 x 10{sup -1} man-Sv, respectively. Thus the order of preference is Baiji, Al-Mahzam, Al-Abbasia, and Abu-Dalaf. The effective dose equivalents to the highest exposed individual resulting from atmospheric as well as aquatic releases of radionuclides from the reactor at any one of the four potential nuclear power-plant sites would not exceed 2 x 10{sup -5} Sv/yr. Thus any one of the four sites is suitable for the operation of the 440 nuclear power plants. 27 refs., 1 tab.

Marouf, B.A.; Al-Kateeb, G.H.; Al-Ani, D.S. [and others

1991-07-01T23:59:59.000Z

178

Prognostics and Health Management in Nuclear Power Plants: A Review of Technologies and Applications  

SciTech Connect (OSTI)

This report reviews the current state of the art of prognostics and health management (PHM) for nuclear power systems and related technology currently applied in field or under development in other technological application areas, as well as key research needs and technical gaps for increased use of PHM in nuclear power systems. The historical approach to monitoring and maintenance in nuclear power plants (NPPs), including the Maintenance Rule for active components and Aging Management Plans for passive components, are reviewed. An outline is given for the technical and economic challenges that make PHM attractive for both legacy plants through Light Water Reactor Sustainability (LWRS) and new plant designs. There is a general introduction to PHM systems for monitoring, fault detection and diagnostics, and prognostics in other, non-nuclear fields. The state of the art for health monitoring in nuclear power systems is reviewed. A discussion of related technologies that support the application of PHM systems in NPPs, including digital instrumentation and control systems, wired and wireless sensor technology, and PHM software architectures is provided. Appropriate codes and standards for PHM are discussed, along with a description of the ongoing work in developing additional necessary standards. Finally, an outline of key research needs and opportunities that must be addressed in order to support the application of PHM in legacy and new NPPs is presented.

Coble, Jamie B.; Ramuhalli, Pradeep; Bond, Leonard J.; Hines, Wes; Upadhyaya, Belle

2012-07-17T23:59:59.000Z

179

Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report  

SciTech Connect (OSTI)

The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.

Ritterbusch, S.E.

2000-08-01T23:59:59.000Z

180

Use of neural networks in the operation of nuclear power plants  

SciTech Connect (OSTI)

Application of neural networks to the operation of nuclear power plants is being investigated under a US Department of Energy sponsored program at the University of Tennessee. Projects include the feasibility of using neural networks for the following tasks: (a) diagnosing specific abnormal conditions, (b) detection of the change of mode of operation, (c) signal validation, (d) monitoring of check valves, (e) modeling of the plant thermodynamics, (f) emulation of core reload calculations, (g) analysis of temporal sequences in NRC's licensee event report,'' (h) monitoring of plant parameters, and (i) analysis of plant vibrations. Each of these projects and its status are described briefly in this article. the objective of each of these projects is to enhance the safety and performance of nuclear plants through the use of neural networks. 6 refs.

Uhrig, R.E. (Tennessee Univ., Knoxville, TN (USA) Oak Ridge National Lab., TN (USA))

1990-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Potential application of neural networks to the operation of nuclear power plants  

SciTech Connect (OSTI)

The application of neural networks, a rapidly evolving technology used extensively in defense applications, to some of the problems of operating nuclear power plants is a logical complement to the expert systems currently being introduced in some of those plants. The potential applications of neural networks include, but are not limited to: (1) Diagnosing specific abnormal conditions. (2) Identifying nonlinear dynamics and transients. (3) Detecting the change of mode of operation. (4) Controlling temperature and pressure during start-up. (5) validating signals. (6) Plant-wide monitoring using autoassociative neural networks. (7) Monitoring of check valves. (8) Modeling the plant thermodynamics to increase efficiency. (9) Emulating core reload calculations. (10) Analyzing temporal sequences in the U.S. Nuclear Regulatory Commission Licensee Event Reports. (11) Monitoring plant parameters. (12) Analyzing vibrations in plants and rotating machinery. The work on such applications indicates that neural networks alone, or in conjunction with other advanced technologies, have the potential to enhance the safety, reliability, and operability of nuclear power plants. 36 refs.

Uhrig, R.E. [University of Tennessee, Knoxville, TN (United States)]|[Oak Ridge National Laboratory, TN (United States)

1991-01-01T23:59:59.000Z

182

The potential role of new technology for enhanced safety and performance of nuclear power plants through improved service maintenance  

E-Print Network [OSTI]

Refinements in the safety and performance of nuclear power plants must be made to maintain public confidence and ensure competitiveness with other power sources. The aircraft industry, US Navy, and other programs have ...

Achorn, Ted Glen

1991-01-01T23:59:59.000Z

183

Nuclear power browning out  

SciTech Connect (OSTI)

When the sad history of nuclear power is written, April 26, 1986, will be recorded as the day the dream died. The explosion at the Chernobyl plant was a terrible human tragedy- and it delivered a stark verdict on the hope that nuclear power will one day replace fossil fuel-based energy systems. Nuclear advocates may soldier on, but a decade after Chernobyl it is clear that nuclear power is no longer a viable energy option for the twenty-first century.

Flavin, C.; Lenssen, N.

1996-05-01T23:59:59.000Z

184

Guideline for the seismic technical evaluation of replacement items for nuclear power plants  

SciTech Connect (OSTI)

Seismic qualification for equipment originally installed in nuclear power plants was typically performed by the original equipment suppliers or manufactures (OES/OEM). Many of the OES/OEM no longer maintain quality assurance programs with adequate controls for supplying nuclear equipment. Utilities themselves must provide reasonable assurance in the continued seismic adequacy of such replacement items. This guideline provides practical, cost-effective techniques which can be used to provide reasonable assurance that replacement items will meet seismic performance requirements necessary to maintain the seismic design basis of commercial nuclear power plants. It also provides a method for determining when a seismic technical evaluation of replacement items (STERI) is required as part of the procurement process for spare and replacement items. Guidance on supplier program requirements necessary to maintain continued seismic adequacy and on documentation of maintaining required seismic adequacy is also included.

Harris, S.P.; Cushing, R.W. (EQE International, San Francisco, CA (United States)); Johnson, H.W. (Programmatic Solutions, Smithtown, NY (United States)); Abeles, J.M. (System 1, Inc., Potomac, MD (United States))

1993-02-01T23:59:59.000Z

185

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information  

SciTech Connect (OSTI)

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

B. Levine

2006-01-27T23:59:59.000Z

186

COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS  

SciTech Connect (OSTI)

This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes “Best Technology Available” for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant’s steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R&D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

Gary Vine

2010-12-01T23:59:59.000Z

187

Aging management of nuclear power plant containments for license renewal  

SciTech Connect (OSTI)

In 1990, the Nuclear Management and Resources Council (NUMARC), now the Nuclear Energy Institute (NEI), submitted for NRC review, the industry reports (IRs), NUMARC Report 90-01 and NUMARC Report 90-10, addressing aging management issues associated with PWR containments and BWR containments for license renewal, respectively. In 1996, the Commission amended 10 CFR 50.55a to promulgate requirements for inservice inspection of containment structures. This rule amendment incorporates by reference the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of the ASME Code addressing the inservice inspection of metal containments/liners and concrete containments, respectively. The purpose of this report is to reconcile the technical information and agreements resulting from the NUMARC IR reviews which are generally described in NUREG-1557 and the inservice inspection requirements of subsections IWE and IWL as promulgated in {section}50.55a for license renewal consideration. This report concludes that Subsections IWE and IWL as endorsed in {section}50.55a are generally consistent with the technical agreements reached during the IR reviews. Specific exceptions are identified and additional evaluations and augmented inspections for renewal are recommended.

Liu, W.C.; Kuo, P.T.; Lee, S.S.

1997-09-01T23:59:59.000Z

188

Devices and methods for managing noncombustible gasses in nuclear power plants  

DOE Patents [OSTI]

Systems passively eliminate noncondensable gasses from facilities susceptible to damage from combustion of built-up noncondensable gasses, such as H2 and O2 in nuclear power plants, without the need for external power and/or moving parts. Systems include catalyst plates installed in a lower header of the Passive Containment Cooling System (PCCS) condenser, a catalyst packing member, and/or a catalyst coating on an interior surface of a condensation tube of the PCCS condenser or an annular outlet of the PCCS condenser. Structures may have surfaces or hydrophobic elements that inhibit water formation and promote contact with the noncondensable gas. Noncondensable gasses in a nuclear power plant are eliminated by installing and using the systems individually or in combination. An operating pressure of the PCCS condenser may be increased to facilitate recombination of noncondensable gasses therein.

Marquino, Wayne; Moen, Stephan C; Wachowiak, Richard M; Gels, John L; Diaz-Quiroz, Jesus; Burns, Jr., John C

2014-12-23T23:59:59.000Z

189

Physical protection solutions for security problems at nuclear power plants. [PWR; BWR  

SciTech Connect (OSTI)

Under Department of Energy sponsorship, Sandia National Laboratories has developed a broad technological base of components and integrated systems to address security concerns at facilities of importance, including nuclear reactors. The primary security concern at a light water reactor is radiological sabotage, a deliberate set of actions at a plant which could expose the public to a significant amount of radiation (on the order of 10 CFR 100 limits). (Also of importance to plant operators are acts of industrial sabotage that could prevent a plant from producing electrical power).

Darby, J.L.; Jacobs, J.

1980-09-01T23:59:59.000Z

190

Extending Sensor Calibration Intervals in Nuclear Power Plants  

SciTech Connect (OSTI)

Currently in the USA, sensor recalibration is required at every refueling outage, and it has emerged as a critical path item for shortening outage duration. International application of calibration monitoring, such as at the Sizewell B plant in UK, has shown that sensors may operate for eight years, or longer, within calibration tolerances. Online monitoring can be employed to identify those sensors which require calibration, allowing for calibration of only those sensors which need it. The US NRC accepted the general concept of online monitoring for sensor calibration monitoring in 2000, but no plants have been granted the necessary license amendment to apply it. This project addresses key issues in advanced recalibration methodologies and provides the science base to enable adoption of best practices for applying online monitoring, resulting in a public domain standardized methodology for sensor calibration interval extension. Research to develop this methodology will focus on three key areas: (1) quantification of uncertainty in modeling techniques used for calibration monitoring, with a particular focus on non-redundant sensor models; (2) accurate determination of acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and (3) the use of virtual sensor estimates to replace identified faulty sensors to extend operation to the next convenient maintenance opportunity.

Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep; Bond, Leonard J.; Shumaker, Brent; Hashemian, Hash

2012-11-15T23:59:59.000Z

191

Aging management guideline for commercial nuclear power plants - tanks and pools  

SciTech Connect (OSTI)

Continued operation of nuclear power plants for periods that extend beyond their original 40-year license period is a desirable option for many U.S. utilities. U.S. Nuclear Regulatory Commission (NRC) approval of operating license renewals is necessary before continued operation becomes a reality. Effective aging management for plant components is important to reliability and safety, regardless of current plant age or extended life expectations. However, the NRC requires that aging evaluations be performed and the effectiveness of aging management programs be demonstrated for components considered within the scope of license renewal before granting approval for operation beyond 40 years. Both the NRC and the utility want assurance that plant components will be highly reliable during both the current license term and throughout the extended operating period. In addition, effective aging management must be demonstrated to support Maintenance Rule (10 CFR 50.65) activities.

Blocker, E.; Smith, S.; Philpot, L.; Conley, J.

1996-02-01T23:59:59.000Z

192

Secretary Bodman Announces Federal Risk Insurance for Nuclear Power Plants  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September2-SCORECARD-01-24-13 Page 1 ofIBRFComplex" at Los

193

DOE Announces Loan Guarantee Applications for Nuclear Power Plant  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power Systems EngineeringDepartmentSmart GridThirdPartnership |Development Center |Construction |

194

Aging of steel containments and liners in nuclear power plants  

SciTech Connect (OSTI)

Aging of the containment pressure boundary in light water reactor plants is being addressed to understand the significant factors relating occurrence of corrosion efficacy of inspection and structural capacity reduction of steel containments and liners of concrete containments. and to make recommendations on use of risk models in regulatory decisions. Current regulatory in-service inspection requirements are reviewed and a summary of containment related degradation experience is presented. Current and emerging nondestructive examination techniques and a degradation assessment methodology for characterizing and quantifying the amount of damage present are described. Quantitative tools for condition assessment of aging structures using time dependent structural reliability analysis methods are summarized. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process. Results of this research provide a means for establishing current and estimating future structural capacity margins of containments, and to address the significance of incidences of reported containment degradation.

Naus, D.J.; Oland, C.B. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.; Ellingwood, B. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering; Norris, W.E. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1998-01-01T23:59:59.000Z

195

Integrated head package cable carrier for a nuclear power plant  

DOE Patents [OSTI]

A cabling arrangement is provided for a nuclear reactor located within a containment. Structure inside the containment is characterized by a wall having a near side surrounding the reactor vessel defining a cavity, an operating deck outside the cavity, a sub-space below the deck and on a far side of the wall spaced from the near side, and an operating area above the deck. The arrangement includes a movable frame supporting a plurality of cables extending through the frame, each connectable at a first end to a head package on the reactor vessel and each having a second end located in the sub-space. The frame is movable, with the cables, between a first position during normal operation of the reactor when the cables are connected to the head package, located outside the sub-space proximate the head package, and a second position during refueling when the cables are disconnected from the head package, located in the sub-space. In a preferred embodiment, the frame straddles the top of the wall in a substantially horizontal orientation in the first position, pivots about an end distal from the head package to a substantially vertically oriented intermediate position, and is guided, while remaining about vertically oriented, along a track in the sub-space to the second position.

Meuschke, Robert E. (Monroeville, PA); Trombola, Daniel M. (Murrysville, PA)

1995-01-01T23:59:59.000Z

196

U.S. Nuclear Power Plant Operating Cost and Experience Summaries  

SciTech Connect (OSTI)

The ''U.S. Nuclear Power Plant Operating Cost and Experience Summaries'' (NUREG/CR-6577, Supp. 2) report has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants during 2000-2001. Costs incurred after initial construction are characterized as annual production costs, which represent fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications, which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operations summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from operating reports submitted by the licensees, the Nuclear Regulatory Commission (NRC) database for enforcement actions, and outage reports.

Reid, RL

2003-09-18T23:59:59.000Z

197

Development of a checklist for evaluating emergency procedures used in nuclear power plants  

SciTech Connect (OSTI)

This report describes the process for developing a checklist to be used by US Nuclear Regulatory Commission Office of Inspection and Enforcement (I and E) inspectors during their evaluation of emergency procedures used in nuclear power plants. The objective of the checklist is to aid inspectors in identifying procedural characteristics that can lead to reactor operator performance deviations. Four nuclear power plants were surveyed to obtain a sample of procedures and related information for human factors evaluation. In addition, a human factors analysis of 890 LERs submitted during the period 1975 through 1978 was performed to identify the major categories of performance deviations associated with reactor operator activities. Checklist items aimed at preventing these performance deviations or facilitating their early detection were developed. The study findings supporting the procedures evaluation criteria comprising the checklist items are described in this report. A companion document, Checklist for Evaluating Emergency Procedures Used in Nuclear Power Plants, NUREG/CR-2005, SAND81-7074, has been prepared as a handbook for inspectors. It describes the checklist and provides instructions for its use. 24 figs.

Brune, R.L.; Weinstein, M.

1981-05-01T23:59:59.000Z

198

Shutdown and low-power operation at commercial nuclear power plants in the United States. Final report  

SciTech Connect (OSTI)

The report contains the results of the NRC Staff`s evaluation of shutdown and low-power operations at US commercial nuclear power plants. The report describes studies conducted by the staff in the following areas: Operating experience related to shutdown and low-power operations, probabilistic risk assessment of shutdown and low-power conditions and utility programs for planning and conducting activities during periods the plant is shut down. The report also documents evaluations of a number of technical issues regarding shutdown and low-power operations performed by the staff, including the principal findings and conclusions. Potential new regulatory requirements are discussed, as well as potential changes in NRC programs. A draft report was issued for comment in February 1992. This report is the final version and includes the responses to the comments along with the staff regulatory analysis of potential new requirements.

Not Available

1993-09-01T23:59:59.000Z

199

Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components  

SciTech Connect (OSTI)

Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four U.S. nuclear steam supply system vendors. For each facility, six locations were studied, including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This report discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants.

Ware, A.G.; Morton, D.K.; Nitzel, M.E. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1995-03-01T23:59:59.000Z

200

Design issues concerning Iran`s Bushehr nuclear power plant VVER-1000 conversion  

SciTech Connect (OSTI)

On January 8, 1995, the Atomic Energy Organization of Iran (AEOI) signed a contract for $800 million with the Russian Federation Ministry for Atomic Energy (Minatom) to complete Bushehr nuclear power plant (BNPP) unit 1. The agreement called for a Russian VVER-1000/320 pressurized water reactor (PWR) to be successfully installed into the existing German-built BNPP facilities in 5 yr. System design differences, bomb damage, and environmental exposure are key issues with which Minatom must contend in order to fulfill the contract. The AEOI under the Shah of Iran envisioned Bushehr as the first of many nuclear power plants, with Iran achieving 24 GW(electric) by 1993 and 34 GW(electric) by 2000. Kraftwerk Union AG (KWU) began construction of the two-unit plant near the Persian Gulf town of Halileh in 1975. Unit 1 was {approx}80% complete and unit 2 was {approx}50% complete when construction was interrupted by the 1979 Iranian Islamic revolution. Despite repeated AEOI attempts to lure KWU and other companies back to Iran to complete the plant, Western concerns about nuclear proliferation in Iran and repeated bombings of the plant during the 1980-1988 Iran-Iraq war dissuaded Germany from resuming construction.

Carson, C.F. [Lawrence Livermore National Laboratory, CA (United States)

1996-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site  

SciTech Connect (OSTI)

This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

L.E. Demick

2011-10-01T23:59:59.000Z

202

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants  

SciTech Connect (OSTI)

This report describes research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

Ramuhalli, Pradeep; Lin, Guang; Crawford, Susan L.; Konomi, Bledar A.; Coble, Jamie B.; Shumaker, Brent; Hashemian, Hash

2014-04-30T23:59:59.000Z

203

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants  

SciTech Connect (OSTI)

This report describes the status of ongoing research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Lin, Guang [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crawford, Susan L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Konomi, Bledar A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Braatz, Brett G. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Coble, Jamie B. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Shumaker, Brent [Analysis and Measurement Services Corp., Knoxville, TN (United States); Hashemian, Hash [Analysis and Measurement Services Corp., Knoxville, TN (United States)

2013-09-01T23:59:59.000Z

204

Managing aging in nuclear power plants: Insights from NRC maintenance team inspection reports  

SciTech Connect (OSTI)

A plant`s maintenance program is the principal vehicle through which age-related degradation is managed. From 1988 to 1991, the NRC evaluated the maintenance program of every nuclear power plant in the United States. Forty-four out of a total of 67 of the reports issued on these in-depth team inspections were reviewed for insights into the strengths and weaknesses of the programs as related to the need to understand and manage the effects of aging on nuclear plant systems, structures, and components. Relevant information was extracted from these inspection reports and sorted into several categories, including Specific Aging Insights, Preventive Maintenance, Predictive Maintenance and Condition Monitoring, Post Maintenance Testing, Failure Trending, Root Cause Analysis and Usage of Probabilistic Risk Assessment in the Maintenance Process. Specific examples of inspection and monitoring techniques successfully used by utilities to detect degradation due to aging have been identified. The information also was sorted according to systems and components, including: Auxiliary Feedwater, Main Feedwater, High Pressure Injection for both BWRs and PWRs, Service Water, Instrument Air, and Emergency Diesel Generator Air Start Systems, and Emergency Diesel Generators Air Start Systems, emergency diesel generators, electrical components such as switchgear, breakers, relays, and motor control centers, motor operated valves and check valves. This information was compared to insights gained from the Nuclear Plant Aging Research (NPAR) Program. Attributes of plant maintenance programs where the NRC inspectors felt that improvement was needed to properly address the aging issue also are discussed.

Fresco, A.; Subudhi, M.; Gunther, W.; Grove, E.; Taylor, J. [Brookhaven National Lab., Upton, NY (United States)

1993-12-01T23:59:59.000Z

205

Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors  

SciTech Connect (OSTI)

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

Not Available

1986-09-01T23:59:59.000Z

206

U.S. Seafood Safe and Unaffected by Radiation Contamination from Japanese Nuclear Power Plant Incident; U.S. Monitoring Control Strategy Explained  

E-Print Network [OSTI]

U.S. Seafood Safe and Unaffected by Radiation Contamination from Japanese Nuclear Power Plant about radiation contamination from the Japanese nuclear power plant incident and on the control potential routes by which seafood contaminated with radionuclides from the Japanese nuclear power plant

207

Redundant Sensor Calibration and Estimation for Monitoring and Control of Nuclear Power Plants Xin Jin, Asok Ray and Robert M. Edwards  

E-Print Network [OSTI]

Redundant Sensor Calibration and Estimation for Monitoring and Control of Nuclear Power Plants Xin@engr.psu.edu INTRODUCTION Performance, reliability and safety of nuclear power plants depend upon validity and accuracy are installed with redundancy in nuclear power plants. Redundancy can be classified into two groups: direct

Ray, Asok

208

The Handbook of Applied Bayesian Analysis, Eds: Tony O'Hagan & Mike West, Oxford University Bayesian analysis and decisions in nuclear power plant  

E-Print Network [OSTI]

Bayesian analysis and decisions in nuclear power plant maintenance Elmira Popova, David Morton, Paul Damien are then applied to solving an important problem in a nuclear power plant system at the South Texas Project (STP) Electric Generation Station. STP is one of the newest and largest nuclear power plants in the US

Morton, David

209

The History and Future of NDE in the Management of Nuclear Power Plant Materials Degradation  

SciTech Connect (OSTI)

The author has spent more than 25 years conducting engineering and research studies to quantify the performance of nondestructive evaluation (NDE) in nuclear power plant (NPP) applications and identifying improvements to codes and standards for NDE to manage materials degradation. This paper will review this fundamental NDE engineering/research work and then look to the future on how NDE can be optimized for proactively managing materials degradation in NPP components.

Doctor, Steven R.

2009-04-01T23:59:59.000Z

210

Dose-projection considerations for emergency conditions at nuclear power plants  

SciTech Connect (OSTI)

The purpose of this report is to review the problems and issues associated with making environmental radiation-dose projections during emergencies at nuclear power plants. The review is divided into three areas: source-term development, characterization of atmospheric dispersion and selection of appropriate dispersion models, and development of dosimetry calculations for determining thyroid dose and whole-body dose for ground-level and elevated releases. A discussion of uncertainties associated with these areas is also provided.

Stoetzel, G.A.; Ramsdell, J.V.; Poeton, R.W.; Powell, D.C.; Desrosiers, A.E.

1983-05-01T23:59:59.000Z

211

Issues for New Nuclear Plants  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

to Explore * Idaho's energy picture * Nuclear power in the U.S. * Potential for a nuclear power plant in Idaho 0 5 10 15 20 25 1960 1970 1980 1990 2000 Million Megawatt-Hours Total...

212

Assessment of the radiological impact of a decommissioning nuclear power plant in Italy  

E-Print Network [OSTI]

The assessment of the radiological impact of a decommissioning Nuclear Power Plant is presented here through the results of an environmental monitoring survey carried out in the area surrounding the Garigliano Power Plant. The levels of radioactivity in soil, water, air and other environmental matrices are shown, in which {\\alpha}, {\\beta} and {\\gamma} activity and {\\gamma} equivalent dose rate are measured. Radioactivity levels of the samples from the Garigliano area are analyzed and then compared to those from a control zone situated more than 100 km away. Moreover, a comparison is made with a previous survey held in 2001. The analyses and comparisons show no significant alteration in the radiological characteristics of the area surroundings the plant, with an overall radioactivity depending mainly from the global fallout and natural sources.

A. Petraglia; C. Sabbarese; M. De Cesare; N. De Cesare; F. Quinto; F. Terrasi; A. D'Onofrio; P. Steier; L. K. Fifield; A. M. Esposito

2012-07-17T23:59:59.000Z

213

Assessment of the radiological impact of a decommissioning nuclear power plant in Italy  

E-Print Network [OSTI]

The assessment of the radiological impact of a decommissioning Nuclear Power Plant is presented here through the results of an environmental monitoring survey carried out in the area surrounding the Garigliano Power Plant. The levels of radioactivity in soil, water, air and other environmental matrices are shown, in which {\\alpha}, {\\beta} and {\\gamma} activity and {\\gamma} equivalent dose rate are measured. Radioactivity levels of the samples from the Garigliano area are analyzed and then compared to those from a control zone situated more than 100 km away. Moreover, a comparison is made with a previous survey held in 2001. The analyses and comparisons show no significant alteration in the radiological characteristics of the area surroundings the plant, with an overall radioactivity depending mainly from the global fallout and natural sources.

Petraglia, A; De Cesare, M; De Cesare, N; Quinto, F; Terrasi, F; D'Onofrio, A; Steier, P; Fifield, L K; Esposito, A M; 10.1051/radiopro/2012010

2012-01-01T23:59:59.000Z

214

Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power  

SciTech Connect (OSTI)

Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

Myers, Carl W [Los Alamos National Laboratory; Elkins, Ned Z [Los Alamos National Laboratory

2008-01-01T23:59:59.000Z

215

Considerations in the evaluation of concrete structures for continued service in aged Nuclear Power Plants (NPPs)  

SciTech Connect (OSTI)

Currently, there are /approximately/119 commercial nuclear power plants (NPPs) in the US either under construction, operating at low-to-full power, or awaiting an operating license. Together, these units have a net generating capacity of /approximately/110 GW(e). Assuming no life extension of present facilities, the operating licenses for these plants will start to expire in the middle of the next decade with Yankee Rowe being the first plant to attain this status. Where it is noted that with no life extension of facilities, a potential loss of electrical generating capacity in excess of 75 GW(e) could occur during the time period 2006 to 2020 when the operating licenses of 80 to 90 NPPs are scheduled to expire. A potential timely and cost-effective solution to meeting future electricity demand, which has worked well for non-nuclear generating plants, is to extend the service life (operating licenses) of existing NPPs. Since the concrete components in these plants provide a vital safety function, any continued service considerations must include an in-depth assessment of the safety-related concrete structures. 7 refs.

Naus, D.; Marchbanks, M.; Oland, B.; Arndt, G.; Brown, T.

1989-01-01T23:59:59.000Z

216

Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants  

SciTech Connect (OSTI)

The Structural Aging Program is addressing the potential for degradation of concrete structural components and systems in nuclear power plants over time due to aging and aggressive environmental stressors. Structures are passive under normal operating conditions but play a key role in mitigating design-basis events, particularly those arising from external challenges such as earthquakes, extreme winds, fires and floods. Structures are plant-specific and unique, often are difficult to inspect, and are virtually impossible to replace. The importance of structural failures in accident mitigation is amplified because such failures may lead to common-cause failures of other components. Structural condition assessment and service life prediction must focus on a few critical components and systems within the plant. Components and systems that are dominant contributors to risk and that require particular attention can be identified through the mathematical formalism of a probabilistic risk assessment, or PRA. To illustrate, the role of structural degradation due to aging on plant risk is examined through the framework of a Level 1 seismic PRA of a nuclear power plant. Plausible mechanisms of structural degradation are found to increase the core damage probability by approximately a factor of two.

Ellingwood, B.; Song, J. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering

1996-03-01T23:59:59.000Z

217

INTERNATIONAL JOURNAL OF HYDROGEN ENERGY Accepted June 2008 HYDROGEN STORAGE FOR MIXED WIND-NUCLEAR POWER PLANTS IN  

E-Print Network [OSTI]

INTERNATIONAL JOURNAL OF HYDROGEN ENERGY Accepted June 2008 1 HYDROGEN STORAGE FOR MIXED WIND-NUCLEAR evaluation of hydrogen production and storage for a mixed wind-nuclear power plant considering some new of a combined nuclear-wind-hydrogen system is discussed first, where the selling and buying of electricity

Cañizares, Claudio A.

218

Decision to reorganise or reorganising decisions? A First-Hand Account of the Decommissioning of the Phnix Nuclear Power Plant  

E-Print Network [OSTI]

of the Decommissioning of the Phénix Nuclear Power Plant Melchior Pelleterat de Borde, MINES ParisTech, Christophe Martin prepared for decommissioning. This study, conducted between 2010 and 2012, is focused on the Phénix nuclear in the context of nuclear decommissioning. This article does not aim to present the results of the study, i

Paris-Sud XI, Université de

219

Aging Management Guideline for commercial nuclear power plants: Battery chargers, inverters and uninterruptible power supplies. Final report  

SciTech Connect (OSTI)

This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant battery chargers, inverters and uninterruptible power supplies important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already, experienced) and aging management program activities to the more generic results and recommendations presented herein.

Berg, R.; Stroinski, M.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

1994-02-01T23:59:59.000Z

220

Numerical simulation of the thermal conditions in a sea bay water area used for water supply to nuclear power plants  

SciTech Connect (OSTI)

Consideration is given to the numerical simulation of the thermal conditions in sea water areas used for both water supply to and dissipation of low-grade heat from a nuclear power plant on the shore of a sea bay.

Sokolov, A. S. [JSC 'B. E. Vedeneev All-Russia Research Institute of Hydraulic Engineering (VNIIG)' (Russian Federation)] [JSC 'B. E. Vedeneev All-Russia Research Institute of Hydraulic Engineering (VNIIG)' (Russian Federation)

2013-07-15T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Risk-informed public safety policy for seismic events in the vicinity of a nuclear power plant  

E-Print Network [OSTI]

Nuclear Power Plants (NPPs) are potentially vulnerable to accidents, which can either be internally or externally initiated. External events include natural events like tornadoes, hurricanes, and earthquakes. The purpose ...

Afolayan Jejeloye, Olubukola

2002-01-01T23:59:59.000Z

222

Comparative analysis of United States and French nuclear power plant siting and construction regulatory policies and their economic consequences  

E-Print Network [OSTI]

Despite the substantial commitments of time and money which are devoted to the nuclear power plant siting process, the effectiveness of the system in providing a balanced evaluation of the technical, environmental and ...

Golay, Michael Warren.

1977-01-01T23:59:59.000Z

223

Development of a hybrid intelligent system for on-line real-time monitoring of nuclear power plant operations  

E-Print Network [OSTI]

A nuclear power plant (NPP) has an intricate operational domain involving systems, structures and components (SSCs) that vary in scale and complexity. Many of the large scale SSCs contribute to the lost availability in the ...

Yildiz, Bilge, 1976-

2003-01-01T23:59:59.000Z

224

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant  

SciTech Connect (OSTI)

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-08-01T23:59:59.000Z

225

Data base on dose reduction research projects for nuclear power plants. Volume 5  

SciTech Connect (OSTI)

This is the fifth volume in a series of reports that provide information on dose reduction research and health physics technology or nuclear power plants. The information is taken from two of several databases maintained by Brookhaven National Laboratory`s ALARA Center for the Nuclear Regulatory Commission. The research section of the report covers dose reduction projects that are in the experimental or developmental phase. It includes topics such as steam generator degradation, decontamination, robotics, improvements in reactor materials, and inspection techniques. The section on health physics technology discusses dose reduction efforts that are in place or in the process of being implemented at nuclear power plants. A total of 105 new or updated projects are described. All project abstracts from this report are available to nuclear industry professionals with access to a fax machine through the ACEFAX system or a computer with a modem and the proper communications software through the ACE system. Detailed descriptions of how to access all the databases electronically are in the appendices of the report.

Khan, T.A.; Yu, C.K.; Roecklein, A.K. [Brookhaven National Lab., Upton, NY (United States)] [Brookhaven National Lab., Upton, NY (United States)

1994-05-01T23:59:59.000Z

226

NUCLEAR PLANT OPERATIONS AND  

E-Print Network [OSTI]

NUCLEAR PLANT OPERATIONS AND CONTROL KEYWORDS: neutron flux, cur- rent noise, vibration diagnostics: Swedish Nuclear Powe

Pázsit, Imre

227

Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants  

SciTech Connect (OSTI)

Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

Kisner, Roger A [ORNL; Mullens, James Allen [ORNL; Wilson, Thomas L [ORNL; Wood, Richard Thomas [ORNL; Korsah, Kofi [ORNL; Qualls, A L [ORNL; Muhlheim, Michael David [ORNL; Holcomb, David Eugene [ORNL; Loebl, Andy [ORNL

2007-08-01T23:59:59.000Z

228

A survey of repair practices for nuclear power plant containment metallic pressure boundaries  

SciTech Connect (OSTI)

The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition.

Oland, C.B.; Naus, D.J. [Oak Ridge National Lab., TN (United States)

1998-05-01T23:59:59.000Z

229

Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program  

SciTech Connect (OSTI)

The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.

Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.

1983-12-01T23:59:59.000Z

230

Method of installing a control room console in a nuclear power plant  

DOE Patents [OSTI]

An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

1994-01-01T23:59:59.000Z

231

Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors  

SciTech Connect (OSTI)

This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55.

Not Available

1985-07-01T23:59:59.000Z

232

Development, Application, and Implementation of RAMCAP to Characterize Nuclear Power Plant Risk From Terrorism  

SciTech Connect (OSTI)

In response to increased interest in risk-informed decision making regarding terrorism, EPRI and ERIN Engineering were selected by U.S. DHS and ASME to develop and demonstrate the RAMCAP method for nuclear power plant (NPP) risk assessment. The objective is to characterize plant-specific NPP risk for risk management opportunities and to provide consistent information for DHS decision making. This paper is an update of this project presented at the American Nuclear Society (ANS) International Topical Meeting on Probabilistic Safety Analysis (PSA05) in September, 2005. The method uses a characterization of risk as a function of Consequence, Vulnerability, and Threat. For each site, worst case scenarios are developed for each of sixteen benchmark threats. Nuclear RAMCAP hypothesizes that the intent of the perpetrator is to cause offsite radiological consequences. Specific targets are the reactor core, the spent fuel pool, and nuclear spent fuel in a dry storage facility (ISFSI). Results for each scenario are presented as conditional risk for financial loss, early fatalities and early injuries. Expected consequences for each scenario are quantified, while vulnerability is estimated on a relative likelihood scale. Insights for other societal risks are provided. Although threat frequencies are not provided, target attractiveness and threat deterrence are estimated. To assure efficiency, completeness, and consistency; results are documented using standard RAMCAP Evaluator software. Trial applications were successfully performed at four plant sites. Implementation at all other U.S. commercial sites is underway, supported by the Nuclear Sector Coordinating Council (NSCC). Insights from RAMCAP results at 23 U.S. plants completed to date have been compiled and presented to the NSCC. Results are site-specific. Physical security barriers, an armed security force, preparedness for design-basis threats, rugged design against natural hazards, multiple barriers between fuel and environment, accident mitigation capability, severe accident management procedures, and offsite emergency plans are risk-beneficial against all threat types. (authors)

Gaertner, John P. [Electric Power Research Institute, 1300 Harris Boulevard, Charlotte, NC 28262 (United States); Teagarden, Grant A. [ERIN Engineering and Research (United States)

2006-07-01T23:59:59.000Z

233

AGE-RELATED DEGRADATION OF NUCLEAR POWER PLANT STRUCTURES AND COMPONENTS.  

SciTech Connect (OSTI)

This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what are the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk.

BRAVERMAN,J.

1999-03-29T23:59:59.000Z

234

Aging management guideline for commercial nuclear power plants-stationary batteries. Final report  

SciTech Connect (OSTI)

The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Berg, R.; Shao, J.; Krencicki, G.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

1994-03-01T23:59:59.000Z

235

Age-Related Degradation of Nuclear Power Plant Structures and Components  

SciTech Connect (OSTI)

This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what was the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk.

Braverman, J.; Chang, T.-Y.; Chokshi, N.; Hofmayer, C.; Morante, R.; Shteyngart, S.

1999-03-29T23:59:59.000Z

236

Aging Management Guideline for commercial nuclear power plants: Electrical switchgear. Final report  

SciTech Connect (OSTI)

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of this AMG to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner which allows personnel responsible for performance analysis and maintenance, to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Toman, G.; Gazdzinski, R.; Schuler, K. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

1993-07-01T23:59:59.000Z

237

ENVIRONMENTAL PROBLEMS ASSOCIATED WITH DECOMMISSIONING THE CHERNOBYL NUCLEAR POWER PLANT COOLING POND  

SciTech Connect (OSTI)

Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.

Farfan, E.

2009-09-30T23:59:59.000Z

238

Environmental Problems Associated With Decommissioning The Chernobyl Nuclear Power Plant Cooling Pond  

SciTech Connect (OSTI)

Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.

Farfan, E. B.; Jannik, G. T.; Marra, J. C.; Oskolkov, B. Ya.; Bondarkov, M. D.; Gaschak, S. P.; Maksymenko, A. M.; Maksymenko, V. M.; Martynenko, V. I.

2009-11-09T23:59:59.000Z

239

Reviewing PSA-based analyses to modify technical specifications at nuclear power plants  

SciTech Connect (OSTI)

Changes to Technical Specifications (TSs) at nuclear power plants (NPPs) require review and approval by the United States Nuclear Regulatory Commission (USNRC). Currently, many requests for changes to TSs use analyses that are based on a plant`s probabilistic safety assessment (PSA). This report presents an approach to reviewing such PSA-based submittals for changes to TSs. We discuss the basic objectives of reviewing a PSA-based submittal to modify NPP TSs; the methodology of reviewing a TS submittal, and the differing roles of a PSA review, a PSA Computer Code review, and a review of a TS submittal. To illustrate this approach, we discuss our review of changes to allowed outage time (AOT) and surveillance test interval (STI) in the TS for the South Texas Project Nuclear Generating Station. Based on this experience gained, a check-list of items is given for future reviewers; it can be used to verify that the submittal contains sufficient information, and also that the review has addressed the relevant issues. Finally, recommended steps in the review process and the expected findings of each step are discussed.

Samanta, P.K.; Martinez-Guridi, G. [Brookhaven National Lab., Upton, NY (United States); Vesely, W.E. [Science Applications International Corporation, Dublin, OH (United States)

1995-12-01T23:59:59.000Z

240

Estimation of the caesium-137 source term from the Fukushima Daiichi nuclear power plant using a consistent joint assimilation of air concentration and  

E-Print Network [OSTI]

plants in Japan. Diesel backup power sys- tems should have sustained the reactors cooling processEstimation of the caesium-137 source term from the Fukushima Daiichi nuclear power plant using during the accident of the Fukushima Daiichi nuclear power plant in March 2011. In Winiarek et al. (2012b

Boyer, Edmond

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1  

SciTech Connect (OSTI)

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

NONE

1995-08-01T23:59:59.000Z

242

Evaluation of aged concrete structures for continued service in nuclear power plants  

SciTech Connect (OSTI)

Results are summarized of a study on concrete component aging and its significance relative to continued service of nuclear power plants (NPPs) beyond the initial period for which they were granted operating licenses. Progress is presented of a second study being conducted to identify and provide acceptance criteria for structural safety issues which the USNRC staff will need to address when applications are submitted for continued service of NPPs. Major activities under this program include: development of a materials property data base, establishment of structural component assessment and repair procedures, and development of a methodology for determination of structural reliability. 19 refs., 5 figs., 3 tabs.

Naus, D.J.; Marchbanks, M.F.; Arndt, E.G.

1988-01-01T23:59:59.000Z

243

Evaluation of aged concrete structures for continued service in nuclear power plants  

SciTech Connect (OSTI)

Results are summarized of a study on concrete component aging and its significance relative to continued service of nuclear power plants (NPPs) beyond the initial period for which they were granted operating licenses. Progress is presented of a second study being conducted to identify and provide acceptance criteria for structural safety issues which the USNRC staff will need to address when applications are submitted for continued service of NPPs. Major activities under this program include: development of a materials property data base, establishment of structural component assessment and repair procedures, and development of a methodology for determination of structural reliability.

Naus, D.J.; Marchbanks, M.F.; Arndt, E.G.

1988-01-01T23:59:59.000Z

244

FRAMEWORK AND APPLICATION FOR MODELING CONTROL ROOM CREW PERFORMANCE AT NUCLEAR POWER PLANTS  

SciTech Connect (OSTI)

This paper summarizes an emerging project regarding the utilization of high-fidelity MIDAS simulations for visualizing and modeling control room crew performance at nuclear power plants. The key envisioned uses for MIDAS-based control room simulations are: (i) the estimation of human error associated with advanced control room equipment and configurations, (ii) the investigative determination of contributory cognitive factors for risk significant scenarios involving control room operating crews, and (iii) the certification of reduced staffing levels in advanced control rooms. It is proposed that MIDAS serves as a key component for the effective modeling of cognition, elements of situation awareness, and risk associated with human performance in next generation control rooms.

Ronald L Boring; David I Gertman; Tuan Q Tran; Brian F Gore

2008-09-01T23:59:59.000Z

245

Proceedings of the Third International Workshop on the implementation of ALARA at nuclear power plants  

SciTech Connect (OSTI)

This report contains the papers presented and the discussions that took place at the Third International Workshop on ALARA Implementation at Nuclear Power Plants, held in Hauppauge, Long Island, New York from May 8--11, 1994. The purpose of the workshop was to bring together scientists, engineers, health physicists, regulators, managers and other persons who are involved with occupational dose control and ALARA issues. The countries represented were: Canada, Finland, France, Germany, Japan, Korea, Mexico, the Netherlands, Spain, Sweden, the United Kingdom and the United States. The workshop was organized into twelve sessions and three panel discussions. Individual papers have been cataloged separately.

Khan, T.A. [comp.] [Brookhaven National Lab., Upton, NY (United States); Roecklein, A.K. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications

1995-03-01T23:59:59.000Z

246

Procedure for conducting a human-reliability analysis for nuclear power plants. Final report  

SciTech Connect (OSTI)

This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the methods described in NUREG/CR-1278, Handbook for Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications. An overview of the procedure describing the major elements of a human reliability analysis is presented along with a detailed description of each element and an example of an actual analysis. An appendix consists of some sample human reliability analysis problems for further study.

Bell, B.J.; Swain, A.D.

1983-05-01T23:59:59.000Z

247

The hunt for theta13 at the Daya Bay nuclear power plant  

E-Print Network [OSTI]

The Daya Bay reactor neutrino experiment is located at the Daya Bay nuclear power plant in Shenzhen, China. The experiment deploys eight "identical" antineutrino detectors to measure antineutrino fluxes from six 2.9 GW_{th} reactor cores in three underground experimental halls at different distances. The target zone of the Daya Bay detector is filled with 20 t 0.1% Gd doped LAB liquid scintillator. The baseline uncorrelated detector uncertainty is ~0.38% using current experimental techniques. Daya Bay can reach a sensitivity of <0.01 to $sin^2 2theta_{13}$ with baseline uncertainties after 3 years of data taking.

Wei Wang; for the Daya Bay collaboration

2009-10-23T23:59:59.000Z

248

The AP1000{sup R} nuclear power plant innovative features for extended station blackout mitigation  

SciTech Connect (OSTI)

Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01T23:59:59.000Z

249

The Decline and Death of Nuclear Power  

E-Print Network [OSTI]

2012). NRC: Nuclear Security and Safeguards.nrc.gov.in nuclear reactor maintenance and security. However, when aof nuclear power plants, as well as physical security to

Melville, Jonathan

2013-01-01T23:59:59.000Z

250

Analysis of a Main Steam Line Break in Asco Nuclear Power Plant  

SciTech Connect (OSTI)

A comprehensive analysis of a double-ended main steam line break (MSLB) accident assumed to occur in the Asco nuclear power plant was carried out using the RELAP/PARCS coupled code. The general results of the benchmark provide a certain qualification of tools and methodologies used. Applying such methodologies to other plant models can be useful to extend conclusions and to identify areas where further analysis is needed. The calculations showed the capability of the control rod to recover the accident. However, one stuck control rod caused some recriticality or return to power (RTP), whose magnitude is heavily affected by the initial and boundary conditions. This paper identifies similarities and discrepancies between the benchmark calculation on the TMI-1 model and the Westinghouse three-loop calculation on the Asco model. The use of an integral plant model was helpful in showing the importance on the RTP of different plant systems that are modeled in detail. The high-pressure injection system and feedwater lines as well as the broken steam line model are the most significant.

Cuadra, Arantxa; Gago, Jose Luis; Reventos, Francesc

2001-06-17T23:59:59.000Z

251

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

average value for nuclear plants) aFinal Envir. Statement (Statement, Koshkonong Nuclear Plant, August 1976. U. S.rem; operation of the nuclear plants themselves only *Other

Nero, A.V.

2010-01-01T23:59:59.000Z

252

Radioactive Releases Impact from Kozloduy Nuclear Power Plant, Bulgaria into the Environment  

SciTech Connect (OSTI)

The aim of this paper is to present a general overview of the radioactive releases impact generated by Kozloduy Nuclear Power Plant (KNPP), Bulgaria to the environment and public. The liquid releases presented are known as the so called controlled water discharges, that are generated after reprocessing of the inevitable accumulated liquid radioactive waste in the plant operation process. The radionuclides containing in the liquid releases are given in the paper as a result of systematic measuring. Database for radiation doses evaluation on the public around Kozloduy NPP site is developed using IAEA LADTAP computerized program. The computer code LADTAP represents realization of a model that evaluates the public dose as a result of NPP releases under normal operation conditions. The results of this evaluation were the basic licensing document for a new liquid release limit.

Genchev, G. T.; Kuleff, I.; Tanev, N. T.; Delistoyanova, E. S.; Guentchev, T.

2002-02-26T23:59:59.000Z

253

Digital control systems in nuclear power plants: Failure information, modeling concepts, and applications. Revision 1  

SciTech Connect (OSTI)

This report briefly describes some current applications of advanced computerized digital display and control systems at US commercial nuclear power plants and presents the results of a literature search that was made to gather information on the reliability of these systems. Both hardware and software reliability were addressed in this review. Only limited failure rate information was found, with the chemical process industry being the primary source of information on hardware failure rates and expert opinion the primary source for software failure rates. Safety-grade digital control systems are typically installed on a functional like-for-like basis, replacing older analog systems without substantially changing interactions with other plant systems. Future work includes performing a limited probabilistic risk assessment of a representative DCS to assess its risk significance.

Galyean, W.J.

1993-06-23T23:59:59.000Z

254

Digital control systems in nuclear power plants: Failure information, modeling concepts, and applications  

SciTech Connect (OSTI)

This report briefly describes some current applications of advanced computerized digital display and control systems at US commercial nuclear power plants and presents the results of a literature search that was made to gather information on the reliability of these systems. Both hardware and software reliability were addressed in this review. Only limited failure rate information was found, with the chemical process industry being the primary source of information on hardware failure rates and expert opinion the primary source for software failure rates. Safety-grade digital control systems are typically installed on a functional like-for-like basis, replacing older analog systems without substantially changing interactions with other plant systems. Future work includes performing a limited probabilistic risk assessment of a representative DCS to assess its risk significance.

Galyean, W.J.

1993-06-23T23:59:59.000Z

255

Handbook of human-reliability analysis with emphasis on nuclear power plant applications. Final report  

SciTech Connect (OSTI)

The primary purpose of the Handbook is to present methods, models, and estimated human error probabilities (HEPs) to enable qualified analysts to make quantitative or qualitative assessments of occurrences of human errors in nuclear power plants (NPPs) that affect the availability or operational reliability of engineered safety features and components. The Handbook is intended to provide much of the modeling and information necessary for the performance of human reliability analysis (HRA) as a part of probabilistic risk assessment (PRA) of NPPs. Although not a design guide, a second purpose of the Handbook is to enable the user to recognize error-likely equipment design, plant policies and practices, written procedures, and other human factors problems so that improvements can be considered. The Handbook provides the methodology to identify and quantify the potential for human error in NPP tasks.

Swain, A D; Guttmann, H E

1983-08-01T23:59:59.000Z

256

Evaluation criteria and procedure for nuclear power plant temporary loads/temporary conditions  

SciTech Connect (OSTI)

Operating nuclear power plants frequently encounter temporary loads/temporary conditions in plant normal operation and maintenance (O and M). The most obvious examples are installation of temporary shielding and scaffolding, or removal of certain supports, to facilitate plant refueling and maintenance outage activities. Short-term operability calls such as those due to snubber failures or unanticipated transients also create temporary loads/temporary conditions. These temporary situations often generate loads that are outside the original plant design basis. Consequently, separate evaluations are needed to ensure that plant structures, systems and components (SSCs) maintain their integrity and functionality while these temporary loads are active. Also, the temporary structures and components need to be evaluated to ensure their integrity during the temporary duration of use. Three types of approaches are normally adopted either individually or in combination to perform needed evaluations: relax the design allowables, use a more refined analysis model but retain the design basis acceptance criteria, or offset temporary loads by eliminating or reducing part of the design basis loads based on short duration considerations. This paper reviews temporary loading/temporary condition issues and the current industry criteria and procedures proposed in dealing with these issues. Where appropriate, regulatory positions on temporary loads/temporary conditions are discussed.

Tang, H.T. [Electric Power Research Inst., Palo Alto, CA (United States); Minichiello, J.C. [Commonwealth Edison Co., Downers Grove, IL (United States); Olson, D.E. [Sargent and Lundy, Chicago, IL (United States)

1996-12-01T23:59:59.000Z

257

A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants  

SciTech Connect (OSTI)

The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

1997-08-01T23:59:59.000Z

258

Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors  

SciTech Connect (OSTI)

The objective of this study was to provide a primer on the environmental effects that can affect the durability of nuclear power plant concrete structures. As concrete ages, changes in its properties will occur as a result of continuing microstructural changes (i.e., slow hydration, crystallization of amorphous constituents, and reactions between cement paste and aggregates), as well as environmental influences. These changes do not have to be detrimental to the point that concrete will not be able to meet its performance requirements. Concrete, however, can suffer undesirable changes with time because of improper specifications, a violation of specifications, or adverse performance of its cement paste matrix or aggregate constituents under either physical or chemical attack. Contained in this report is a discussion on concrete durability and the relationship between durability and performance, a review of the historical perspective related to concrete and longevity, a description of the basic materials that comprise reinforced concrete, and information on the environmental factors that can affect the performance of nuclear power plant concrete structures. Commentary is provided on the importance of an aging management program.

Naus, Dan J [ORNL

2007-02-01T23:59:59.000Z

259

Monitoring Thermal Fatigue Damage In Nuclear Power Plant Materials Using Acoustic Emission  

SciTech Connect (OSTI)

Proactive aging management of nuclear power plant passive components requires technologies to enable monitoring and accurate quantification of material condition at early stages of degradation (i.e., pre-macrocrack). Acoustic emission (AE) is well-suited to continuous monitoring of component degradation and is proposed as a method to monitor degradation during accelerated thermal fatigue tests. A key consideration is the ability to separate degradation responses from external sources such as water spray induced during thermal fatigue testing. Water spray provides a significant background of acoustic signals, which can overwhelm AE signals caused by degradation. Analysis of AE signal frequency and energy is proposed in this work as a means for separating degradation signals from background sources. Encouraging results were obtained by applying both frequency and energy filters to preliminary data. The analysis of signals filtered using frequency and energy provides signatures exhibiting several characteristics that are consistent with degradation accumulation in materials. Future work is planned to enable verification of the efficacy of AE for thermal fatigue crack initiation detection. While the emphasis has been placed on the use of AE for crack initiation detection during accelerated aging tests, this work also has implications with respect to the use of AE as a primary tool for early degradation monitoring in nuclear power plant materials. The development of NDE tools for characterization of aging in materials can also benefit from the use of a technology such as AE which can continuously monitor and detect crack initiation during accelerated aging tests.

Meyer, Ryan M.; Ramuhalli, Pradeep; Watson, Bruce E.; Pitman, Stan G.; Roosendaal, Timothy J.; Bond, Leonard J.

2012-04-26T23:59:59.000Z

260

The evolution of the break preclusion concept for nuclear power plants in Germany  

SciTech Connect (OSTI)

In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A = open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.

Schulz, H. [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

1997-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Challenges in Determining the Isotopic Mixture for the Fukushima Daiichi Nuclear Power Plant  

SciTech Connect (OSTI)

As part of the United States response to the Fukushima Daiichi Nuclear Power Plant emergency, the National Nuclear Security Administration (NNSA) Consequence Management (CM) Teams were activated with elements deploying to Japan. The NNSA CM teams faced the urgent need for information regarding the potential radiological doses that citizens of might experience. This paper discusses the challenges and lessons learned associated with the analysis of field collected samples and gamma spectra in an attempt to determine the isotopic mixture present on the ground around the Plant. There were several interesting and surprising lessons to be learned from the sample analysis portion of the response. The paper discusses several elements of the response that were unique to the event occurring in Japan, as well as several elements that would have occurred in a U.S. nuclear reactor event. Sections of this paper address details of the specific analytical challenges faced during the efforts to analyze samples and try to understand the overall release source term.

Shanks, Arthur [Sandia National Laboratories; Fournier, Sean [Sandia National Laboratories; Shanks, Sonoya [Sandia National Laboratories

2012-05-01T23:59:59.000Z

262

A practical approach to risk-based inservice inspection in U.S. nuclear power plants  

SciTech Connect (OSTI)

To provide guidelines for practical implementation of risk-based ISI, EPRI sponsored work to develop evaluation procedures and criteria for defining risk-based inservice inspection programs for nuclear power plant piping. These procedures and criteria include efficient means to identify risk significant piping segments, inspection locations, and available inspection techniques. These procedures were applied in a pilot study to assess the feasibility of successfully implementing risk-based inservice inspection programs at nuclear plants. The results from the pilot study indicate that implementation of risk-based inservice inspection programs can reduce the cost and radiation exposure associated with inservice inspection, while maintaining a high level of safety. The list of references provides additional details of these procedures and plant-specific applications. Also, an EPRI technical report has been published to document these procedures. Software has been developed to support and fully document this procedure. Additional development is adding an expert system to the present data base system. The approach compares well to approaches used (or being considered) in other industries and can easily be adapted to these other industries and to address economic and personnel safety in addition to public safety measures.

Gosselin, S.R. [Electric Power Research Inst., Charlotte, NC (United States); Gamble, R. [Sartrex Corp., Rockville, MD (United States); Dimitrijevic, V.B.; O`Regan, P.J.; Chapman, J.R. [Yankee Atomic Electric Co., Bolton, MS (United States)

1996-12-01T23:59:59.000Z

263

Recommendations for NRC policy on shift scheduling and overtime at nuclear power plants  

SciTech Connect (OSTI)

This report contains the Pacific Northwest Laboratory's (PNL's) recommendations to the US Nuclear Regulatory Commission (NRC) for an NRC policy on shift scheduling and hours of work (including overtime) for control room operators and other safety-related personnel in nuclear power plants. First, it is recommended that NRC make three additions to its present policy on overtime: (1) limit personnel to 112 hours of work in a 14-day period, 192 hours in 28 days, and 2260 hours in one year; exceeding these limits would require plant manager approval; (2) add a requirement that licensees obtain approval from NRC if plant personnel are expected to exceed 72 hours of work in a 7-day period, 132 hours in 14 days, 228 hours in 28 days, and 2300 hours in one year; and (3) make the policy a requirement, rather than a nonbinding recommendation. Second, it is recommended that licensees be required to obtain NRC approval to adopt a routine 12-hour/day shift schedule. Third, it is recommended that NRC add several nonbinding recommendations concerning routine 8-hour/day schedules. Finally, because additional data can strengthen the basis for future NRC policy on overtime, five methods are suggested for collecting data on overtime and its effects. 44 refs., 10 tabs.

Lewis, P.M.

1985-07-01T23:59:59.000Z

264

The Decline and Death of Nuclear Power  

E-Print Network [OSTI]

The Economist (2012). Nuclear power: The 30-year itch. Thesince the Cold War, nuclear power plants are being plannedDramatic fall in new nuclear power stations after Fukushima.

Melville, Jonathan

2013-01-01T23:59:59.000Z

265

Formation of hot particles during the Chernobyl nuclear power plant accident  

SciTech Connect (OSTI)

The oxidation of irradiated Chernobyl nuclear fuel at 670 to 1,170 K for 3 to 21 h resulted in its destruction into fine particles, the dispersal composition of which is well described by lognormal distribution regularity. The median radius of the formed particles does not depend on the annealing temperature and decreases with the increase of the annealing period from 10 to 3 {micro}m. Proceeding from the dispersal composition and matrix composition of the Chernobyl hot fuel particles, it can be concluded that the oxidation of nuclear fuel was one of the basic mechanisms of hot fuel particle formation during the accident at the Chernobyl nuclear power plant. With oxidation in air and the dispersal of irradiated oxide nuclear fuel at as low as 670 K, ruthenium, located on the granular borders, is released. Ruthenium is oxidized to volatile RuO{sub 4}, sublimated, and condensed on materials of iron. Nickel and stainless steel can be efficiently used at high temperatures (tested to 1,200 K) for radioruthenium adsorption in accidents and for some technological operations. As the temperature of hot fuel particles annealed in inert media increases from 1,270 to 2,270 K, the relative release of radionuclides increases in the following sequence: cesium isotopes; europium isotopes; cerium isotopes; americium isotopes; and ruthenium, plutonium, and curium isotopes.

Kashparov, V.A.; Ivanov, Y.A.; Zvarisch, S.I.; Protsak, V.P.; Khomutinin, Y.V.; Kurepin, A.D.; Pazukhin, E.M. [Ukrainian Inst. of Agricultural Radiology, Chabany (Ukraine)

1996-05-01T23:59:59.000Z

266

An Integrated Scheme for Anomaly Identification and Automatic Control of Nuclear Power Plants Xin Jin, Robert M. Edwards and Asok Ray  

E-Print Network [OSTI]

An Integrated Scheme for Anomaly Identification and Automatic Control of Nuclear Power Plants Xin.edu INTRODUCTION Nuclear Power Plants (NPPs) are complex systems with many variables that require adjustment Jin, Robert M. Edwards and Asok Ray Department of Mechanical and Nuclear Engineering, The Pennsylvania

Ray, Asok

267

U.S. nuclear power plants as terrorist targets : threat perception and the media  

E-Print Network [OSTI]

In recent history, nuclear engineers and the nuclear power industry have been primarily concerned with two things: safety and waste. In the past few years, a third concern has risen to join these two at the top: terrorism. ...

Laughter, Mark, 1980-

2005-01-01T23:59:59.000Z

268

A Review of Sensor Calibration Monitoring for Calibration Interval Extension in Nuclear Power Plants  

SciTech Connect (OSTI)

Currently in the United States, periodic sensor recalibration is required for all safety-related sensors, typically occurring at every refueling outage, and it has emerged as a critical path item for shortening outage duration in some plants. Online monitoring can be employed to identify those sensors that require calibration, allowing for calibration of only those sensors that need it. International application of calibration monitoring, such as at the Sizewell B plant in United Kingdom, has shown that sensors may operate for eight years, or longer, within calibration tolerances. This issue is expected to also be important as the United States looks to the next generation of reactor designs (such as small modular reactors and advanced concepts), given the anticipated longer refueling cycles, proposed advanced sensors, and digital instrumentation and control systems. The U.S. Nuclear Regulatory Commission (NRC) accepted the general concept of online monitoring for sensor calibration monitoring in 2000, but no U.S. plants have been granted the necessary license amendment to apply it. This report presents a state-of-the-art assessment of online calibration monitoring in the nuclear power industry, including sensors, calibration practice, and online monitoring algorithms. This assessment identifies key research needs and gaps that prohibit integration of the NRC-approved online calibration monitoring system in the U.S. nuclear industry. Several needs are identified, including the quantification of uncertainty in online calibration assessment; accurate determination of calibration acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and assessment of the feasibility of using virtual sensor estimates to replace identified faulty sensors in order to extend operation to the next convenient maintenance opportunity. Understanding the degradation of sensors and the impact of this degradation on signals is key to developing technical basis to support acceptance criteria and set point decisions, particularly for advanced sensors which do not yet have a cumulative history of operating performance.

Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep; Bond, Leonard J.; Hashemian, Hash; Shumaker, Brent; Cummins, Dara

2012-08-31T23:59:59.000Z

269

Aging assessment of essential HVAC chillers used in nuclear power plants  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory conducted a comprehensive aging assessment of chillers used in the essential safety air-conditioning systems in nuclear power plants (NPPs). The chillers used, and air-conditioning systems served, vary in design from plant to plant. The review of operating experience indicated that chillers experience aging degradation and failures. The primary aging factors of concern for chillers include vibration, excessive temperatures and pressures, thermal cycling, chemical attack, and poor quality cooling water. The evaluation of Licensee Event Reports (LERs) indicated that about 38% of the failures were primarily related to aging, 55% were partially aging related, and 7% of the failures were unassignable. About 25% of the failures were primarily caused by human, design, procedure, and other errors. The large number of errors is probably directly related to the complexity of chillers and their interfacing systems. Nearly all of the LERs were the result of entering plant Technical Specification Limiting Condition for Operation (LCO) that initiated remedial actions like plant shutdown procedures. The trend for chiller-related LERs has stabilized at about 0.13 LERs per plant year since 1988. Carefully following the vendor procedures and monitoring the equipment can help to minimize and/or eliminate most of the premature failures. Recording equipment performance can be useful for trending analysis. Periodic operation for a few hours on a weekly or monthly basis is useful to remove moisture and non-condensable gases that gradually build up inside the chiller. Chiller pressurization kits are available that will help minimize the amount of moisture and air ingress to low-pressure chillers during standby periods. The assessment of service life condition monitoring of chillers indicated there are many simple to sophisticated methods available that can help in chiller surveillance and monitoring.

Blahnik, D.E.; Camp, T.W.

1996-09-01T23:59:59.000Z

270

Applying Human Factors Evaluation and Design Guidance to a Nuclear Power Plant Digital Control System  

SciTech Connect (OSTI)

The United States (U.S.) nuclear industry, like similar process control industries, has moved toward upgrading its control rooms. The upgraded control rooms typically feature digital control system (DCS) displays embedded in the panels. These displays gather information from the system and represent that information on a single display surface. In this manner, the DCS combines many previously separate analog indicators and controls into a single digital display, whereby the operators can toggle between multiple windows to monitor and control different aspects of the plant. The design of the DCS depends on the function of the system it monitors, but revolves around presenting the information most germane to an operator at any point in time. DCSs require a carefully designed human system interface. This report centers on redesigning existing DCS displays for an example chemical volume control system (CVCS) at a U.S. nuclear power plant. The crucial nature of the CVCS, which controls coolant levels and boration in the primary system, requires a thorough human factors evaluation of its supporting DCS. The initial digital controls being developed for the DCSs tend to directly mimic the former analog controls. There are, however, unique operator interactions with a digital vs. analog interface, and the differences have not always been carefully factored in the translation of an analog interface to a replacement DCS. To ensure safety, efficiency, and usability of the emerging DCSs, a human factors usability evaluation was conducted on a CVCS DCS currently being used and refined at an existing U.S. nuclear power plant. Subject matter experts from process control engineering, software development, and human factors evaluated the DCS displays to document potential usability issues and propose design recommendations. The evaluation yielded 167 potential usability issues with the DCS. These issues should not be considered operator performance problems but rather opportunities identified by experts to improve upon the design of the DCS. A set of nine design recommendations was developed to address these potential issues. The design principles addressed the following areas: (1) color, (2) pop-up window structure, (3) navigation, (4) alarms, (5) process control diagram, (6) gestalt grouping, (7) typography, (8) terminology, and (9) data entry. Visuals illustrating the improved DCS displays accompany the design recommendations. These nine design principles serve as the starting point to a planned general DCS style guide that can be used across the U.S. nuclear industry to aid in the future design of effective DCS interfaces.

Thomas Ulrich; Ronald Boring; William Phoenix; Emily Dehority; Tim Whiting; Jonathan Morrell; Rhett Backstrom

2012-08-01T23:59:59.000Z

271

Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants  

SciTech Connect (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.

O, J.M.; Higgins, J.; Stephen Fleger - NRC

2011-09-19T23:59:59.000Z

272

Behavior-based rules for fitness-for-duty assessment of nuclear power plant personnel  

SciTech Connect (OSTI)

The safe and reliable operation of nuclear power plants requires that plant personnel not be under the influence of any substance, legal or illegal, or mentally or physically impaired from any cause that in any way adversely affects their ability to safely and competently perform their duties. This goal has been formalized by the US Nuclear Regulatory Commission in their proposed rule for a fitness-for-duty program. The purpose of this paper is to describe a performance-based tool based on surrogate tests and dose equivalency methodologies that is a viable candidate for fitness-for-duty assessment. The automated performance test system (APTS) is a microcomputer-based human performance test battery that has been developed over a decade of research supported variously by the National Science Foundation, National Aeronautics and Space Administration, US Department of Energy, and the US Navy and Army. Representing the most psychometrically sound test from evaluations of over 150 well-known tests of basic psychomotor and cognitive skills, the battery provides direct prediction of a worker's fitness for duty. Twenty-four tests are suitable for use, and a dozen have thus far been shown to be sensitive to the effects of legal and illegal drugs, alcohol, fatigue, stress, and other causes of impairment.

Kennedy, R.S.; Turnage, J.J.; Price, H.E.; Lane, N.E.

1989-01-01T23:59:59.000Z

273

REVIEW Of COMPUTERIZED PROCEDURE GUIDELINES FOR NUCLEAR POWER PLANT CONTROL ROOMS  

SciTech Connect (OSTI)

Computerized procedures (CPs) are recognized as an emerging alternative to paper-based procedures for supporting control room operators in nuclear power plants undergoing life extension and in the concept of operations for advanced reactor designs. CPs potentially reduce operator workload, yield increases in efficiency, and provide for greater resilience. Yet, CPs may also adversely impact human and plant performance if not designed and implemented properly. Therefore, it is important to ensure that existing guidance is sufficient to provide for proper implementation and monitoring of CPs. In this paper, human performance issues were identified based on a review of the behavioral science literature, research on computerized procedures in nuclear and other industries, and a review of industry experience with CPs. The review of human performance issues led to the identification of a number of technical gaps in available guidance sources. To address some of the gaps, we developed 13 supplemental guidelines to support design and safety. This paper presents these guidelines and the case for further research.

David I Gertman; Katya Le Blanc; Ronald L Boring

2011-09-01T23:59:59.000Z

274

Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR  

SciTech Connect (OSTI)

This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

1983-08-01T23:59:59.000Z

275

Digital Full-Scope Simulation of a Conventional Nuclear Power Plant Control Room, Phase 2: Installation of a Reconfigurable Simulator to Support Nuclear Plant Sustainability  

SciTech Connect (OSTI)

The U.S. Department of Energy’s Light Water Reactor Sustainability program has developed a control room simulator in support of control room modernization at nuclear power plants in the U.S. This report highlights the recent completion of this reconfigurable, full-scale, full-scope control room simulator buildout at the Idaho National Laboratory. The simulator is fully reconfigurable, meaning it supports multiple plant models developed by different simulator vendors. The simulator is full-scale, using glasstop virtual panels to display the analog control boards found at current plants. The present installation features 15 glasstop panels, uniquely achieving a complete control room representation. The simulator is also full-scope, meaning it uses the same plant models used for training simulators at actual plants. Unlike in the plant training simulators, the deployment on glasstop panels allows a high degree of customization of the panels, allowing the simulator to be used for research on the design of new digital control systems for control room modernization. This report includes separate sections discussing the glasstop panels, their layout to mimic control rooms at actual plants, technical details on creating a multi-plant and multi-vendor reconfigurable simulator, and current efforts to support control room modernization at U.S. utilities. The glasstop simulator provides an ideal testbed for prototyping and validating new control room concepts. Equally importantly, it is helping create a standardized and vetted human factors engineering process that can be used across the nuclear industry to ensure control room upgrades maintain and even improve current reliability and safety.

Ronald L. Boring; Vivek Agarwal; Kirk Fitzgerald; Jacques Hugo; Bruce Hallbert

2013-03-01T23:59:59.000Z

276

Aging assessment of essential HVAC chillers used in nuclear power plants. Phase 1, Volume 1  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory conducted a Phase I aging assessment of chillers used in the essential safety air-conditioning systems of nuclear power plants. Centrifugal chillers in the 75- to 750-ton refrigeration capacity range are the predominant type used. The chillers used, and air-conditioning systems served, vary in design from plant-to-plant. It is crucial to keep chiller internals very clean and to prevent the leakage of water, air, and other contaminants into the refrigerant containment system. Periodic operation on a weekly or monthly basis is necessary to remove moisture and noncondensable gases that gradually build up inside the chiller. This is especially desirable if a chiller is required to operate only as an emergency standby unit. The primary stressors and aging mechanisms that affect chillers include vibration, excessive temperatures and pressures, thermal cycling, chemical attack, and poor quality cooling water. Aging is accelerated by moisture, non-condensable gases (e.g., air), dirt, and other contamination within the refrigerant containment system, excessive start/stop cycling, and operating below the rated capacity. Aging is also accelerated by corrosion and fouling of the condenser and evaporator tubes. The principal cause of chiller failures is lack of adequate monitoring. Lack of performing scheduled maintenance and human errors also contribute to failures.

Blahnik, D.E.; Klein, R.F. [Pacific Northwest Lab., Richland, WA (United States)

1993-09-01T23:59:59.000Z

277

aguirre nuclear plant: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

278

Emptying of the Storage for Solid Radioactive Waste in the Greifswald Nuclear Power Plant  

SciTech Connect (OSTI)

On the Greifswald site, 8 WWER 440 reactor units are located and also several facilities to handle fuel and radwaste. After the reunification of Germany, the final decision was taken to decommission all these Russian designed reactors. Thus, EWN is faced with a major decommissioning project in the field of nuclear power stations. One of the major tasks before the dismantling of the plant is the complete disposal of the operational waste. Among other facilities, a store for solid radioactive waste is located on the site, which has been filled over 17 years of operation of units 1 to 4. The paper presents the disposal technology development and results achieved. This activity is the first project in the operational history of the Russian type serial reactor line WWER-440.

Hartmann, B.; Fischer, J.

2002-02-26T23:59:59.000Z

279

Industrial Complex for Solid Radwaste Management at Chernobyle Nuclear Power Plant  

SciTech Connect (OSTI)

In the framework of the preparation for the decommissioning of the Chernobyl Nuclear Power Plant (ChNPP) an Industrial Complex for Solid Radwaste Management (ICSRM) will be built under the EC TACIS Program in the vicinity of ChNPP. The paper will present the proposed concepts and their integration into existing buildings and installations. Further, the paper will consider the safety cases, as well as the integration of Western and Ukrainian Organizations into a cohesive project team and the requirement to guarantee the fulfillment of both Western standards and Ukrainian regulations and licensing requirements. The paper will provide information on the status of the interim design and the effects of value engineering on the output of basic design phase. The paper therefor summarizes the design results of the involved design engineers of the Design and Process Providers BNFL (LOT 1), RWE NUKEM GmbH (LOT 2 and General) and INITEC (LOT 3).

Ahner, S.; Fomin, V. V.

2002-02-26T23:59:59.000Z

280

Review of Methods Related to Assessing Human Performance in Nuclear Power Plant Control Room Simulations  

SciTech Connect (OSTI)

With the increased use of digital systems in Nuclear Power Plant (NPP) control rooms comes a need to thoroughly understand the human performance issues associated with digital systems. A common way to evaluate human performance is to test operators and crews in NPP control room simulators. However, it is often challenging to characterize human performance in meaningful ways when measuring performance in NPP control room simulations. A review of the literature in NPP simulator studies reveals a variety of ways to measure human performance in NPP control room simulations including direct observation, automated computer logging, recordings from physiological equipment, self-report techniques, protocol analysis and structured debriefs, and application of model-based evaluation. These methods and the particular measures used are summarized and evaluated.

Katya L Le Blanc; Ronald L Boring; David I Gertman

2001-11-01T23:59:59.000Z

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to obtain the most current and comprehensive results.


281

Integrated Diagnostic and Prognostic Tools for Residual Life Estimation in Aging Nuclear Power Plant Components  

SciTech Connect (OSTI)

Recent events in Japan have focused renewed attention on the safe operation of light water reactor (LWR) nuclear power plants (NPPs). A central issue in safe, long-term operations of existing and planned NPPs is the early detection and monitoring of significant materials degradation. Materials aging and degradation in passive components is expected to be the key factor in determining the operational life of an NPP and may limit long-term operations in the current LWR fleet. Methods for detecting and assessing the degradation state in NPP structural materials, followed by approaches to estimate the remaining useful life (RUL) of the component, are therefore necessary for safe, long-term operations. This paper explores advanced diagnostic and prognostic approaches to detecting material degradation, and then determining RUL given the current material state.

Ramuhalli, Pradeep; Meyer, Ryan M.; Bond, Leonard J.; Griffin, Jeffrey W.; Henager, Charles H.

2011-06-01T23:59:59.000Z

282

Human-centered HMI design to support cognitive process of operators in nuclear power plants  

SciTech Connect (OSTI)

In this study, an operation advisory system to aid cognitive process of operators is proposed for advanced main control rooms (MCRs) in future nuclear power plants (NPPs). As MCRs are fully digitalized and designed based on computer technologies, MCRs have much evolved by improving human-machine interface (HMI) design and by adapting automation or support systems for helping operator's convenient operation and maintenance. Various kinds of support systems for operators are developed or developing for advanced MCRs. The proposed system is suggesting a design basis about 'What kinds of support systems are most efficient and necessary for MCR operators ' and 'how to use them together.' In this paper, the operator's operation processes are analyzed based on a human cognitive process model and appropriate support systems that support each activity of the human cognitive process are suggested. Also, the proposed support system is evaluated using Bayesian belief network model and human error probabilities in order to estimate its effect. (authors)

Lee, S. J.; Seong, P. H. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Inst. of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2006-07-01T23:59:59.000Z

283

Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR  

SciTech Connect (OSTI)

The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.

Morris, F.A.; Hooper, R.L.

1983-07-01T23:59:59.000Z

284

Detection of $^{133}$Xe from the Fukushima nuclear power plant in the upper troposphere above Germany  

E-Print Network [OSTI]

After the accident in the Japanese Fukushima Dai-ichi nuclear power plant in March 2011 large amounts of radioactivity were released and distributed in the atmosphere. Among them were also radioactive noble gas isotopes which can be used as tracers to probe global atmospheric circulation models. This work presents unique measurements of the radionuclide $^{133}$Xe from Fukushima in the upper troposphere above Germany. The measurements involve air sampling in a research jet aircraft followed by chromatographic xenon extraction and ultra-low background gas counting with miniaturized proportional counters. With this technique a detection limit of the order of 100 $^{133}$Xe atoms in liter-scale air samples (corresponding to about 100 mBq/m$^3$) is achievable. Our results proof that the $^{133}$Xe-rich ground level air layer from Fukushima was lifted up to the tropopause and distributed hemispherically. Moreover, comparisons with ground level air measurements indicate that the arrival of the radioactive plume in ...

Simgen, Hardy; Aufmhoff, Heinfried; Baumann, Robert; Kaether, Florian; Lindemann, Sebastian; Rauch, Ludwig; Schlager, Hans; Schlosser, Clemens; Schumann, Ulrich

2013-01-01T23:59:59.000Z

285

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink  

SciTech Connect (OSTI)

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, just can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I and C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results. (authors)

Meng Lin; Dong Hou; Zhihong Xu; Yanhua Yang; Ronghua Zhang [Shanghai Jiao Tong University, Shanghai, 200030 (China)

2006-07-01T23:59:59.000Z

286

RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

for Fossil-Fu.e l and Geothermal Power Plants", Lawrencefrom fossil-fuel and geothermal power plants Control offrom fossil-fuel and geothermal power plants Radionuclide

Nero, A.V.

2010-01-01T23:59:59.000Z

287

atomic power plants: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

288

atomic power plant: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

289

aagesta nuclear power: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

290

OECD/NEA study on the economics of the long-term operation of nuclear power plants  

SciTech Connect (OSTI)

The OECD Nuclear Energy Agency (NEA) established the Ad hoc expert group on the Economics of Long-term Operation (LTO) of Nuclear Power Plants. The primary aim of this group is to collect and analyse technical and economic data on the upgrade and lifetime extension experience in OECD countries, and to assess the likely applications for future extensions. This paper describes the key elements of the methodology of economic assessment of LTO and initial findings for selected NEA member countries. (authors)

Lokhov, A.; Cameron, R. [OECD Nuclear Energy Agency, 12, boulevard des Iles, 92130 Issy-les-Moulineaux (France)

2012-07-01T23:59:59.000Z

291

Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings  

SciTech Connect (OSTI)

ABSTRACT Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 and US NRC Standard Review Plan, Section 3.8) ; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 and 10 CFR 50); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 as well as SECY 90-016, SECY 93-087, and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

Braverman, J.I.; Morante, R.

2010-07-18T23:59:59.000Z

292

Comparative Evaluation of Cutting Methods of Activated Concrete from Nuclear Power Plant Decommissioning - 13548  

SciTech Connect (OSTI)

The amount of radioactive wastes from decommissioning of a nuclear power plant varies greatly depending on factors such as type and size of the plant, operation history, decommissioning options, and waste treatment and volume reduction methods. There are many methods to decrease the amount of decommissioning radioactive wastes including minimization of waste generation, waste reclassification through decontamination and cutting methods to remove the contaminated areas. According to OECD/NEA, it is known that the radioactive waste treatment and disposal cost accounts for about 40 percentage of the total decommissioning cost. In Korea, it is needed to reduce amount of decommissioning radioactive waste due to high disposal cost, about $7,000 (as of 2010) per a 200 liter drum for the low- and intermediate-level radioactive waste (LILW). In this paper, cutting methods to minimize the radioactive waste of activated concrete were investigated and associated decommissioning cost impact was assessed. The cutting methods considered are cylindrical and volume reductive cuttings. The study showed that the volume reductive cutting is more cost-effective than the cylindrical cutting. Therefore, the volume reductive cutting method can be effectively applied to the activated bio-shield concrete. (authors)

Kim, HakSoo; Chung, SungHwan; Maeng, SungJun [Central Research Institute, Korea Hydro and Nuclear Power Co. Ltd., 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)] [Central Research Institute, Korea Hydro and Nuclear Power Co. Ltd., 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

2013-07-01T23:59:59.000Z

293

Agricultural approaches of remediation in the outside of the Fukushima Daiichi nuclear power plant  

SciTech Connect (OSTI)

This paper outlines agricultural approaches of remediation activity done in contaminated areas around the Fukushima Daiichi Nuclear Power Plant. About the decontamination examination of contaminated areas, we have tried the land scale test of a rice field before and after planting by the use of currently recommended methods. Since farmers would carry out the land preparation by themselves, generation of secondary radioactive waste should be as low as possible through the decontamination works. For the radioactive nuclide migration control of rice by wet rice production, several types of decontamination methods such as zeolite addition and potassium fertilization in the soil have been examined. The results are summarized in the 4 following points. 1) Plowing and water discharge are effective for removing radioactive cesium from rice field. 2) Additional potassium fertilization is effective for reducing cesium radioactivity in the product. 3) No significant difference is observed with or without the zeolite addition. 4) Very low transfer factor of cesium from soil to brown rice has been obtained compared with literature values.

Sato, Nobuaki [Tohoku University, 2-1-1 Katahira Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Saso, Michitaka [Toshiba Corporation Power Systems Company: 2-1 Ukishima-cho, Kawasaki-ku, Kawasaki, Kanagawa 210-0862 (Japan); Umeda, Miki [Japan Atomic Energy Agency, 4-29 Muramatsu, Tokai, Ibaraki 319-1184 (Japan); Fujii, Yasuhiko [Tokyo Institute of Technology:2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Amemiya, Kiyoshi [Hazama Corporation: 2-2-5 Toranomon, Minato-ku, Tokyo 105-8479 (Japan)

2013-07-01T23:59:59.000Z

294

Nuclear power high technology colloquium: proceedings  

SciTech Connect (OSTI)

Reports presenting information on technology advancements in the nuclear industry and nuclear power plant functions have been abstracted and are available on the energy data base.

Not Available

1984-12-10T23:59:59.000Z

295

An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing  

SciTech Connect (OSTI)

The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T. [Sandia National Labs., Albuquerque, NM (United States)

1996-05-01T23:59:59.000Z

296

The trend of digital control system design for nuclear power plants in Korea  

SciTech Connect (OSTI)

Currently there are 20 nuclear power plants (NPPs) in operation, and 6 more units are under construction in Korea. The control systems of those NPPs have also been developed together with the technology advancement. Control systems started with On-Off control using the relay logic, had been evolved into Solid-State logic using TTL ICs, and applied with the micro-processors since the Yonggwang NPP Units 3 and 4 which started its construction in 1989. Multiplexers are also installed at the local plant areas to collect field input and to send output signals while communicating with the controllers located in the system cabinets near the main control room in order to reduce the field wiring cables. The design of the digital control system technology for the NPPs in Korea has been optimized to maximize the operability as well as the safety through the design, construction, start-up and operation experiences. Both Shin-Kori Units 1 and 2 and Shin-Wolsong Units 1 and 2 NPP projects under construction are being progressed at the same time. Digital Plant Control Systems of these projects have adopted multi-loop controllers, redundant loop configuration, and soft control system for the radwaste system. Programmable Logic Controller (PLC) and Distributed Control System (DCS) are applied with soft control system in Shin-Kori Units 3 and 4. This paper describes the evolvement of control system at the NPPs in Korea and the experience and design improvement through the observation of the latest failure of the digital control system. In addition, design concept and its trend of the digital control system being applied to the NPP in Korea are introduced. (authors)

Park, S. H.; Jung, H. Y.; Yang, C. Y.; Choe, I. N. [Korea Power Engineering Company, 360-9 Mabuk-Dong, Yongin-Si, Gyeonggi-Do, 446-713 (Korea, Republic of)

2006-07-01T23:59:59.000Z

297

Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants  

SciTech Connect (OSTI)

This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

1998-01-01T23:59:59.000Z

298

Evaluation of station blackout accidents at nuclear power plants: Technical findings related to unresolved safety issue A-44: Final report  

SciTech Connect (OSTI)

''Station Blackout,'' which is the complete loss of alternating current (AC) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on AC power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the fequency of loss of offsite power; the probability that emergency or onsite AC power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of AC power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without AC power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events.

Not Available

1988-06-01T23:59:59.000Z

299

Review of nuclear power plant safety cable aging studies with recommendations for improved approaches and for future work.  

SciTech Connect (OSTI)

Many U. S. nuclear power plants are approaching 40 years of age and there is a desire to extend their life for up to 100 total years. Safety-related cables were originally qualified for nuclear power plant applications based on IEEE Standards that were published in 1974. The qualifications involved procedures to simulate 40 years of life under ambient power plant aging conditions followed by simulated loss of coolant accident (LOCA). Over the past 35 years or so, substantial efforts were devoted to determining whether the aging assumptions allowed by the original IEEE Standards could be improved upon. These studies led to better accelerated aging methods so that more confident 40-year lifetime predictions became available. Since there is now a desire to potentially extend the life of nuclear power plants way beyond the original 40 year life, there is an interest in reviewing and critiquing the current state-of-the-art in simulating cable aging. These are two of the goals of this report where the discussion is concentrated on the progress made over the past 15 years or so and highlights the most thorough and careful published studies. An additional goal of the report is to suggest work that might prove helpful in answering some of the questions and dealing with some of the issues that still remain with respect to simulating the aging and predicting the lifetimes of safety-related cable materials.

Gillen, Kenneth Todd; Bernstein, Robert

2010-11-01T23:59:59.000Z

300

Threatened and endangered species evaluation for 75 licensed commercial nuclear power generating plants  

SciTech Connect (OSTI)

The Endangered Species Act (ESA) of 1973, as amended, and related implementing regulations of the jurisdictional federal agencies, the U.S. Departments of Commerce and Interior, at 50 CFR Part 17. 1, et seq., require that federal agencies ensure that any action authorized, funded, or carried out under their jurisdiction is not likely to jeopardize the continued existence of any threatened or endangered species or result in the destruction or adverse modification of critical habitats for such species. The issuance and maintenance of a federal license, such as a construction permit or operating license issued by the U.S. Nuclear Regulatory Commission (NRC) for a commercial nuclear power generating facility is a federal action under the jurisdiction of a federal agency, and is therefore subject to the provisions of the ESA. The U.S. Department of the Interior (through the Fish and Wildlife Service), and the U.S. Department of Commerce, share responsibility for administration of the ESA. The National Marine Fisheries Service (NMFS) deals with species that inhabit marine environments and anadromous fish, while the U.S. Fish and Wildlife Service (USFWS) is responsible for terrestrial and freshwater species and migratory birds. A species (or other distinct taxonomic unit such as subspecies, variety, and for vertebrates, distinct population units) may be classified for protection as `endangered` when it is in danger of extinction within the foreseeable future throughout all or a significant portion of its range. A `threatened` classification is provided to those animals and plants likely to become endangered within the foreseeable future throughout all or a significant portion of their ranges. As of February 1997, there were about 1067 species listed under the ESA in the United States. Additionally there were approximately 125 species currently proposed for listing as threatened or endangered, and another 183 species considered to be candidates for formal listing proposals.

Sackschewsky, M.R.

1997-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

An Empirical Study on Ultrasonic Testing in Lieu of Radiography for Nuclear Power Plants  

SciTech Connect (OSTI)

Research is being conducted for the U.S. Nuclear Regulatory Commission (NRC) at the Pacific Northwest National Laboratory (PNNL) to assess the capability, effectiveness, and reliability of ultrasonic testing (UT) as a replacement method for radiographic testing (RT) for inspecting nuclear power plant (NPP) components. A primary objective of this work is to evaluate UT techniques to assess their ability to detect, locate, size, and characterize fabrication flaws in typical NPP weldments. This particular study focused on the evaluation of four carbon steel pipe-to-pipe welds on specimens that ranged in thicknesses from 19.05 mm (0.75 in.) to 27.8 mm (1.094 in.) and were 355.6 mm (14.0 in.) or 406.4 mm (16.0 in.) in diameter. The pipe welds contained both implanted (intentional) fabrication flaws as well as bonus (unintentional) flaws throughout the entire thickness of the weld and the adjacent base material. The fabrication flaws were a combination of planar and volumetric flaw types, including incomplete fusion, incomplete penetration, cracks, porosity, and slag inclusions. The examinations were conducted using phased-array UT (PA UT) techniques applied primarily for detection and length sizing of the flaws. Radiographic examinations were also conducted on the specimens with RT detection and length sizing results being used to establish true state. This paper will discuss the comparison of UT and RT (true state) detection results conducted to date along with a discussion on the technical gaps that need to be addressed before these methods can be used interchangeably for repair and replacement activities for NPP components.

Moran, Traci L.; Pardini, Allan F.; Ramuhalli, Pradeep; Prowant, Matthew S.; Mathews, Royce

2012-09-01T23:59:59.000Z

302

Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant  

SciTech Connect (OSTI)

The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories are used to perform scaled experiments that simulate High Pressure Melt Ejection accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt (thermite) is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic air/steam/hydrogen atmospheres, and hydrogen generation and combustion, can be studied. Four Integral Effects Tests (IETs) have been performed with scale models of the Surry NPP to investigate DCH phenomena. The 1/61{sup th} scale Integral Effects Tests (IET-9, IET-10, and IET-11) were conducted in CTRF, which is a 1/6{sup th} scale model of the Surry reactor containment building (RCB). The 1/10{sup th} scale IET test (IET-12) was performed in the Surtsey vessel, which had been configured as a 1/10{sup th} scale Surry RCB. Scale models were constructed in each of the facilities of the Surry structures, including the reactor pressure vessel, reactor support skirt, control rod drive missile shield, biological shield wall, cavity, instrument tunnel, residual heat removal platform and heat exchangers, seal table room and seal table, operating deck, and crane wall. This report describes these experiments and gives the results.

Blanchat, T.K.; Allen, M.D.; Pilch, M.M. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

1994-06-01T23:59:59.000Z

303

Progress and Status of the Ignalina Nuclear Power Plant's New Solid Waste Management and Storage Facilities  

SciTech Connect (OSTI)

A considerable amount of dry radioactive waste from former NPP operation has accumulated up to date and is presently stored at the Ignalina NPP site, Lithuania. Current storage capacities are nearly exhausted and more waste is to come from future decommissioning of the two RMBKtype reactors. Additionally, the existing storage facilities does not comply to the state-of-the-art technology for handling and storage of radioactive waste. In 2005, INPP faced this situation of a need for waste processing and subsequent interim storage of these wastes by contracting NUKEM with the design, construction, installation and commissioning of new waste management and storage facilities. The subject of this paper is to describe the scope and the status of the new solid waste management and storage facilities at the Ignalina Nuclear Power Plant. In summary: The turnkey contract for the design, supply and commission of the SWMSF was awarded in December 2005. The realisation of the project was initially planned within 48 month. The basic design was finished in August 2007 and the Technical Design Documentation and Preliminary Safety Analyses Report was provided to Authorities in October 2007. The construction license is expected in July 2008. The procurement phase was started in August 2007, start of onsite activities is expected in November 2007. The start of operation of the SWMSF is scheduled for end of 2009. (authors)

Rausch, J.; Henderson, R.W. [NUKEM Technologies GmbH, Alzenau (Germany); Penkov, V. [State Enterprise Ignalina Nuclear Power Plant, Visaginas (Lithuania)

2008-07-01T23:59:59.000Z

304

Detection of $^{133}$Xe from the Fukushima nuclear power plant in the upper troposphere above Germany  

E-Print Network [OSTI]

After the accident in the Japanese Fukushima Dai-ichi nuclear power plant in March 2011 large amounts of radioactivity were released and distributed in the atmosphere. Among them were also radioactive noble gas isotopes which can be used as tracers to test global atmospheric circulation models. This work presents unique measurements of the radionuclide $^{133}$Xe from Fukushima in the upper troposphere above Germany. The measurements involve air sampling in a research jet aircraft followed by chromatographic xenon extraction and ultra-low background gas counting with miniaturized proportional counters. With this technique a detection limit of the order of 100 $^{133}$Xe atoms in litre-scale air samples (corresponding to about 100 mBq/m$^3$) is achievable. Our results provide proof that the $^{133}$Xe-rich ground level air layer from Fukushima was lifted up to the tropopause and distributed hemispherically. Moreover, comparisons with ground level air measurements indicate that the arrival of the radioactive plume at high altitude over Germany occurred several days before the ground level plume.

Hardy Simgen; Frank Arnold; Heinfried Aufmhoff; Robert Baumann; Florian Kaether; Sebastian Lindemann; Ludwig Rauch; Hans Schlager; Clemens Schlosser; Ulrich Schumann

2014-12-05T23:59:59.000Z

305

EIS No. 20100312 EIS Comanche Peak Nuclear Power Plant Units 3 and 4  

SciTech Connect (OSTI)

In accordance with Section 309(a) of the Clean Air Act, EPA is required to make its comments on EISs issued by other Federal agencies public. Historically, EPA has met this mandate by publishing weekly notices of availability of EPA comments, which includes a brief summary of EPA's comment letters, in the Federal Register. Since February 2008, EPA has been including its comment letters on EISs on its Web site at: http://www.epa.gov/compliance/nepa/eisdata.html. Including the entire EIS comment letters on the Web site satisfies the Section 309(a) requirement to make EPA's comments on EISs available to the public. Accordingly, on March 31, 2010, EPA discontinued the publication of the notice of availability of EPA comments in the Federal Register. EIS No. 20100312, Draft EIS, NRC, TX, Comanche Peak Nuclear Power Plant Units 3 and 4, Application for Combined Licenses (COLs) for Construction Permits and Operating Licenses, (NUREG-1943), Hood and Somervell Counties, TX, Comment Period Ends: 10/26/2010.

Bjornstad, David J [ORNL

2010-08-01T23:59:59.000Z

306

Future Prospects for Nuclear Power after Fukushima  

E-Print Network [OSTI]

at the FukushimaDaiichi nuclear power plant in Japan has changed the perception of nuclear as a safe energy sourceFuture Prospects for Nuclear Power after Fukushima Nuclear is a highintensity energy source as the next generation of Light Water Reactors. We will also discuss the future prospects of nuclear power

Goldberg, Bennett

307

COMPUTER-BASED PROCEDURE FOR FIELD ACTIVITIES: RESULTS FROM THREE EVALUATIONS AT NUCLEAR POWER PLANTS  

SciTech Connect (OSTI)

Nearly all activities that involve human interaction with the systems of a nuclear power plant are guided by procedures. The paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety; however, improving procedure use could yield tremendous savings in increased efficiency and safety. One potential way to improve procedure-based activities is through the use of computer-based procedures (CBPs). Computer-based procedures provide the opportunity to incorporate context driven job aids, such as drawings, photos, just-in-time training, etc into CBP system. One obvious advantage of this capability is reducing the time spent tracking down the applicable documentation. Additionally, human performance tools can be integrated in the CBP system in such way that helps the worker focus on the task rather than the tools. Some tools can be completely incorporated into the CBP system, such as pre-job briefs, placekeeping, correct component verification, and peer checks. Other tools can be partly integrated in a fashion that reduces the time and labor required, such as concurrent and independent verification. Another benefit of CBPs compared to PBPs is dynamic procedure presentation. PBPs are static documents which limits the degree to which the information presented can be tailored to the task and conditions when the procedure is executed. The CBP system could be configured to display only the relevant steps based on operating mode, plant status, and the task at hand. A dynamic presentation of the procedure (also known as context-sensitive procedures) will guide the user down the path of relevant steps based on the current conditions. This feature will reduce the user’s workload and inherently reduce the risk of incorrectly marking a step as not applicable and the risk of incorrectly performing a step that should be marked as not applicable. As part of the Department of Energy’s (DOE) Light Water Reactors Sustainability Program, researchers at Idaho National Laboratory (INL) along with partners from the nuclear industry have been investigating the design requirements for computer-based work instructions (including operations procedures, work orders, maintenance procedures, etc.) to increase efficiency, safety, and cost competitiveness of existing light water reactors.

Oxstrand, Johanna [Idaho National Laboratory; Bly, Aaron [Idaho National Laboratory; LeBlanc, Katya [Idaho National Laboratory

2014-09-01T23:59:59.000Z

308

POWER PLANT RELIABILITY-AVAILABILITY AND STATE REGULATION. VOLUME 7 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

nuclear, geothermal, and fossil-fuel power plants. However,power plants, which are reviewed and licensed by the Nuclear Regulatory Commission (NRC), and relatively few areas of geothermal and

Nero, A.V.

2010-01-01T23:59:59.000Z

309

Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants  

SciTech Connect (OSTI)

OAK-B135 This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies.

Camillo A. DiNunzio Framatome ANP DE& S; Dr. Abhinav Gupta Assistant Professor NCSU; Dr. Michael Golay Professor MIT Dr. Vincent Luk Sandia National Laboratories; Rich Turk Westinghouse Electric Company Nuclear Systems; Charles Morrow, Sandia National Laboratories; Geum-Taek Jin, Korea Power Engineering Company Inc.

2002-11-30T23:59:59.000Z

310

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants  

SciTech Connect (OSTI)

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

Woo, H.H.; Lu, S.C.

1981-09-15T23:59:59.000Z

311

Climate Change, Nuclear Power and Nuclear  

E-Print Network [OSTI]

Climate Change, Nuclear Power and Nuclear Proliferation: Magnitude Matters Rob Goldston MIT IAP plays a large role in replacing coal red plants. al hydro electricity options penetrate in the climate way across scenarios, showing a slight severe climate targets. In Industry, the climate target has

312

Seismic fragility evaluation of a piping system in a nuclear power plant by shaking table test and numerical analysis  

SciTech Connect (OSTI)

In this study, a seismic fragility evaluation of the piping system in a nuclear power plant was performed. For the evaluation of seismic fragility of the piping system, this research was progressed as three steps. At first, several piping element capacity tests were performed. The monotonic and cyclic loading tests were conducted under the same internal pressure level of actual nuclear power plants to evaluate the performance. The cracks and wall thinning were considered as degradation factors of the piping system. Second, a shaking tale test was performed for an evaluation of seismic capacity of a selected piping system. The multi-support seismic excitation was performed for the considering a difference of an elevation of support. Finally, a numerical analysis was performed for the assessment of seismic fragility of piping system. As a result, a seismic fragility for piping system of NPP in Korea by using a shaking table test and numerical analysis. (authors)

Kim, M. K.; Kim, J. H.; Choi, I. K. [Korea Atomic Energy Research Inst., Daedeok-daero 989-111, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2012-07-01T23:59:59.000Z

313

Features of adsorbed radioactive chemical elements and their isotopes distribution in iodine air filters AU-1500 at nuclear power plants  

E-Print Network [OSTI]

The main aim of research is to investigate the physical features of spatial distribution of the adsorbed radioactive chemical elements and their isotopes in the granular filtering medium in the iodine air filters of the type of AU1500 in the forced exhaust ventilation systems at the nuclear power plant. The gamma activation analysis method is applied to accurately characterize the distribution of the adsorbed radioactive chemical elements and their isotopes in the granular filtering medium in the AU1500 iodine air filter after its long term operation at the nuclear power plant. The typical spectrum of the detected chemical elements and their isotopes in the AU1500 iodine air filter, which was exposed to the bremsstrahlung gamma quantum irradiation, produced by the accelerating electrons in the tantalum target, are obtained. The spatial distributions of the detected chemical element 127I and some other chemical elements and their isotopes in the layer of absorber, which was made of the cylindrical coal granule...

Neklyudov, I M; Dikiy, N P; Ledenyov, O P; Lyashko, Yu V

2013-01-01T23:59:59.000Z

314

Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant  

SciTech Connect (OSTI)

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant.

Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States); Garner, L.W. [Nuclear Regulatory Commission, Washington, DC (United States)

1993-08-01T23:59:59.000Z

315

Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant  

SciTech Connect (OSTI)

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant.

Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

1994-05-01T23:59:59.000Z

316

Auxiliary feedwater system risk-based inspection guide for the J. M. Farley Nuclear Power Plant  

SciTech Connect (OSTI)

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment(PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab.

Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G. (Pacific Northwest Lab., Richland, WA (USA))

1990-10-01T23:59:59.000Z

317

Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant  

SciTech Connect (OSTI)

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab.

Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E. (Pacific Northwest Lab., Richland, WA (United States))

1991-09-01T23:59:59.000Z

318

Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant  

SciTech Connect (OSTI)

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant.

Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

1993-12-01T23:59:59.000Z

319

Security during the Construction of New Nuclear Power Plants: Technical Basis for Access Authorization and Fitness-For-Duty Requirements  

SciTech Connect (OSTI)

A technical letter report to the NRC summarizing the findings of a benchmarking study, literature review, and workshop with experts on current industry standards and expert judgments about needs for security during the construction phase of critical infrastructure facilities in the post-September 11 U.S. context, with a special focus on the construction phase of nuclear power plants and personnel security measures.

Branch, Kristi M.; Baker, Kathryn A.

2009-09-01T23:59:59.000Z

320

A best estimate method for the diagnosis and mitigation of multiple-failure transients in nuclear power plants  

E-Print Network [OSTI]

, he only monitors one source, such as the pressurizer pressure indicator, rather than examining other instruments that can supply the same information, such as reactor vessel pressure or the pump outlet pressure. This procedure of monitoring on only..., concentrating on applying a themy of probability, However, the work presented herein is based on a different method involving "confidence levels". For this project a qualitative model of a nuclear power plant and the ATMS knowledge base of transient facts...

Martin, Robert Paul

2012-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Radioactivity pollution and protection of underground waters within the location of nuclear power plants in Jaslovske Bohunice  

SciTech Connect (OSTI)

As a result of research conducted at the Nuclear Power Plant (NPP) A-1 in connection with the decommissioning of the A-1 reactor, tritium contamination has been found in the ground water. A program has been undertaken for the monitoring and protection of underground waters, both onsite and offsite. The paper describes the present level of knowledge on the actual hydrogeological and radiological status of the area.

Plsko, J.; Kostolansky, M. [EKOSUR, Trnava (Slovakia); Polak, R. [HYDROPOL, Bratislava (Slovakia)

1993-12-31T23:59:59.000Z

322

Surfactants containing radioactive run-offs: Ozone treatment, influence on nuclear power plants water waste special treatment  

SciTech Connect (OSTI)

The authors discuss the problems encountered in the efficiency of radioactive waste treatment in nuclear power plants in Kursk. The ozonization of aqueous solutions of surfactants was carried out in the laboratory`s ozonization system. The surfactants which are discharged to the ion exchangers deteriorate resins, clog up the ion exchangers, and decrease filtration velocity. Therefore, this investigation focused on finding a method to increase the efficiency of this treatment process.

Prokudina, S.A.; Grachok, M.A. [Belarussian State Economic Univ., Minsk (Belarus)

1993-12-31T23:59:59.000Z

323

Regulatory analysis for amendments to regulations for the environmental review for renewal of nuclear power plant operating licenses. Final report  

SciTech Connect (OSTI)

This regulatory analysis provides the supporting information for a proposed rule that will amend the Nuclear Regulatory Commission`s environmental review requirements for applications for renewal of nuclear power plant operating licenses. The objective of the proposed rulemaking is to improve regulatory efficiency by providing for the generic evaluation of certain environmental impacts associated with nuclear plant license renewal. After considering various options, the staff identified and analyzed two major alternatives. With Alternative A, the existing regulations would not be amended. This option requires that environmental reviews be performed under the existing regulations. Alternative B is to assess, on a generic basis, the environmental impacts of renewing the operating license of individual nuclear power plants, and define the issues that will need to be further analyzed on a case-by-case basis. In addition, Alternative B removes from NRC`s review certain economics-related issues. The findings of this assessment are to be codified in 10 CFR 51. The staff has selected Alternative B as the preferred alternative.

NONE

1996-05-01T23:59:59.000Z

324

Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion  

SciTech Connect (OSTI)

This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

Sayre, Edwin D. [Engineering Consultant, 218 Brooke Acres Drive, Los Gatos, CA 95032 (United States); Ring, Peter J. [Advanced Methods and Materials, 1190 Mountain View-Alviso Rd. Suite P, Sunnyvale, CA 94089 (United States); Brown, Neil [Engineering Consultant, 5134 Cordoy Lane, San Jose, CA 95124 (United States); Elsner, Norbert B.; Bass, John C. [Hi-Z Technology, Inc., 7606 Miramar Rd. Suite 7400, San Diego, CA 92126 (United States)

2008-01-21T23:59:59.000Z

325

Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America  

SciTech Connect (OSTI)

The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

2011-11-01T23:59:59.000Z

326

Issues arising with the application of optical fiber transmission in class 1E systems in nuclear power plants  

SciTech Connect (OSTI)

The application of fiber optic links and networks in safety-critical systems in the next generation of nuclear power plants, as well as in some digital upgrades in present-day plants, will mean that these links must be highly reliable and able to withstand the effect of environmental stressors present at the installation location. This paper discusses the failure modes and age-related mechanisms of fiber optic transmission components and identifies environmental stressors that could adversely affect their reliability over the long term. Some of the standards that could be used in their qualification for safety-critical applications are also discussed briefly.

Korsah, K. [Oak Ridge National Lab., TN (United States); Antonescu, C. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1993-12-31T23:59:59.000Z

327

Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants  

SciTech Connect (OSTI)

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab.

Moffitt, N.E.; Gore, B.F.: Vo, T.V. (Pacific Northwest Lab., Richland, WA (USA))

1991-07-01T23:59:59.000Z

328

As officials in Japan deal with the accumulation of radioactive seawater near the devastated Fukushima Daiichi nuclear power plant in the wake of last month's  

E-Print Network [OSTI]

Fukushima Daiichi nuclear power plant in the wake of last month's earthquake and tsunami, the U.S. Department of Energy is investing in fundamental research it hopes can be used to build safer nuclear reactors and avoid reactor emergencies. The department's Nuclear Criticality Safety Program (NCSP

Danon, Yaron

329

Simulation of operational transients in a VVER-1000 nuclear power plant using the RELAP5/MOD3.2 computer program  

E-Print Network [OSTI]

A RELAP5/MOD3.2 nodalization model of a VVER-1OOO (V-320) nuclear power plant was updated, improved and validated against available experimental data. The data included integrated test results obtained from actual power plant testing. The steady...

Moscalu, Dionisie Radu

1999-01-01T23:59:59.000Z

330

RADIOACTIVE WASTE MANAGEMENT IN THE CHERNOBYL EXCLUSION ZONE - 25 YEARS SINCE THE CHERNOBYL NUCLEAR POWER PLANT ACCIDENT  

SciTech Connect (OSTI)

Radioactive waste management is an important component of the Chernobyl Nuclear Power Plant accident mitigation and remediation activities of the so-called Chernobyl Exclusion Zone. This article describes the localization and characteristics of the radioactive waste present in the Chernobyl Exclusion Zone and summarizes the pathways and strategy for handling the radioactive waste related problems in Ukraine and the Chernobyl Exclusion Zone, and in particular, the pathways and strategies stipulated by the National Radioactive Waste Management Program. The brief overview of the radioactive waste issues in the ChEZ presented in this article demonstrates that management of radioactive waste resulting from a beyond-designbasis accident at a nuclear power plant becomes the most challenging and the costliest effort during the mitigation and remediation activities. The costs of these activities are so high that the provision of radioactive waste final disposal facilities compliant with existing radiation safety requirements becomes an intolerable burden for the current generation of a single country, Ukraine. The nuclear accident at the Fukushima-1 NPP strongly indicates that accidents at nuclear sites may occur in any, even in a most technologically advanced country, and the Chernobyl experience shows that the scope of the radioactive waste management activities associated with the mitigation of such accidents may exceed the capabilities of a single country. Development of a special international program for broad international cooperation in accident related radioactive waste management activities is required to handle these issues. It would also be reasonable to consider establishment of a dedicated international fund for mitigation of accidents at nuclear sites, specifically, for handling radioactive waste problems in the ChEZ. The experience of handling Chernobyl radioactive waste management issues, including large volumes of radioactive soils and complex structures of fuel containing materials can be fairly useful for the entire world's nuclear community and can help make nuclear energy safer.

Farfan, E.; Jannik, T.

2011-10-01T23:59:59.000Z

331

Initial quantification of human error associated with specific instrumentation and control system components in licensed nuclear power plants  

SciTech Connect (OSTI)

This report provides a methodology for the initial quantification of specific categories of human errors made in conjunction with several instrumentation and control (I and C) system components operated, maintained, and tested in licensed nuclear power plants. The resultant human error rates (HER) provide the first real systems bases of comparison for the existing derived and/or best judgement equivalent set of such rates or probabilities. These calculated error rates also provide the first real indication of human performance as it relates directly to specific tasks in nuclear plants. This work of developing specific HERs is both an extension of and an outgrowth of the generic HERs developed for safety system pumpc and valves as reported in NUREG/CR-1880.

Luckas, W.J. Jr.; Lettieri, V.; Hall, R.E.

1982-02-01T23:59:59.000Z

332

Initial quantification of human error associated with specific instrumentation and control system components in licensed nuclear power plants  

SciTech Connect (OSTI)

This report provides a methodology for the initial quantification of specific categories of human errors made in conjunction with several instrumentation and control (I and C) system components operated, maintained, and tested in licensed nuclear power plants. The resultant human error rates (HER) provide the first real systems bases of comparison for the existing derived and/or best judgement equivalent set of such rates or probabilities. These calculated error rates also provide the first real indication of human performance as it relates directly to specific tasks in nuclear plants. This work of developing specific HERs is both an extension of and an outgrowth of the generic HERs developed for safety system pumps and valves as reported in NUREG/CR-1880.

Luckas, W.J. Jr.; Lettieri, V.; Hall, R.E.

1982-02-01T23:59:59.000Z

333

E-Print Network 3.0 - aged nuclear power Sample Search Results  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

4 Background Analog electro-mechanical systems in existing nuclear power plants are aging... Engineering 2 Nuclear Power Plant 12;2 MIT Department of Nuclear Engineering 3...

334

A Procedure for Determination of Degradation Acceptance Criteria for Structures and Passive Components in Nuclear Power Plants  

SciTech Connect (OSTI)

The Korea Atomic Energy Research Institute (KAERI) has been collaborating with Brookhaven National Laboratory since 2007 to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). This collaboration program aims at providing technical support to a five-year KAERI research project, which includes three specific areas that are essential to seismic probabilistic risk assessment: (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. The understanding and assessment of age-related degradations of structures, systems, and components and their impact on plant safety is the major goal of this KAERI-BNL collaboration. Four annual reports have been published before this report as a result of the collaboration research.

Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y-S.; Hahm, D.; Choi, I-K.

2012-01-30T23:59:59.000Z

335

Aging of turbine drives for safety-related pumps in nuclear power plants  

SciTech Connect (OSTI)

This study was performed to examine the relationship between time-dependent degradation and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety-related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized-water reactor plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented.

Cox, D.F. [Oak Ridge National Lab., TN (United States)

1995-06-01T23:59:59.000Z

336

NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944  

E-Print Network [OSTI]

#12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20, naval reactors, and nuclear power plants. Oak Ridge experiments byArt Snell in 1944 showed that 10 tons

Pennycook, Steve

337

CEC-150-2006-001-F NUCLEAR POWER  

E-Print Network [OSTI]

on California's nuclear power plants and key nuclear power issues such as nuclear waste storage, disposal, and transportation. The report reviews the federal and state regulatory framework for nuclear power and the various of continuing to operate California's aging nuclear power plants. Safety and security issues are key

338

Power Plant Cycling Costs  

SciTech Connect (OSTI)

This report provides a detailed review of the most up to date data available on power plant cycling costs. The primary objective of this report is to increase awareness of power plant cycling cost, the use of these costs in renewable integration studies and to stimulate debate between policymakers, system dispatchers, plant personnel and power utilities.

Kumar, N.; Besuner, P.; Lefton, S.; Agan, D.; Hilleman, D.

2012-07-01T23:59:59.000Z

339

Nuclear power plant Generic Aging Lessons Learned (GALL). Main report and appendix A  

SciTech Connect (OSTI)

The purpose of this generic aging lessons learned (GALL) review is to provide a systematic review of plant aging information in order to assess materials and component aging issues related to continued operation and license renewal of operating reactors. Literature on mechanical, structural, and thermal-hydraulic components and systems reviewed consisted of 97 Nuclear Plant Aging Research (NPAR) reports, 23 NRC Generic Letters, 154 Information Notices, 29 Licensee Event Reports (LERs), 4 Bulletins, and 9 Nuclear Management and Resources Council Industry Reports (NUMARC IRs) and literature on electrical components and systems reviewed consisted of 66 NPAR reports, 8 NRC Generic Letters, 111 Information Notices, 53 LERs, 1 Bulletin, and 1 NUMARC IR. More than 550 documents were reviewed. The results of these reviews were systematized using a standardized GALL tabular format and standardized definitions of aging-related degradation mechanisms and effects. The tables are included in volume s 1 and 2 of this report. A computerized data base has also been developed for all review tables and can be used to expedite the search for desired information on structures, components, and relevant aging effects. A survey of the GALL tables reveals that all ongoing significant component aging issues are currently being addressed by the regulatory process. However, the aging of what are termed passive components has been highlighted for continued scrutiny. This document is Volume 1, consisting of the executive summary, summary and observations, and an appendix listing the GALL literature review tables.

Kaza, K.E.; Diercks, D.R.; Holland, J.W.; Choi, S.U. [and others

1996-12-01T23:59:59.000Z

340

The (safety-related) heat exchangers aging management guideline for commercial nuclear power plants, and developments since 1994  

SciTech Connect (OSTI)

The US Department of Energy (DOE), in cooperation with the Electric Power Research Institute (EPRI) and US nuclear power plant utilities, is preparing a series of aging management guidelines (AMGs) for commodity types of components (e.g., heat exchangers, electrical cable and terminations, pumps). Commodities are included in this series based on their importance to continued nuclear plant operation and license renewal. The AMGs contain a detailed summary of operating history, stressors, aging mechanisms, and various types of maintenance and surveillance practices that can be combined to create an effective aging management program. Each AMG is intended for use by the systems engineers and plant maintenance staff (i.e., an AMG is intended to be a hands-on technical document rather than a licensing document). The heat exchangers AMG, published in June 1994, includes the following information of interest to nondestructive examination (NDE) personnel: aging mechanisms determined to be non-significant for all applications; aging mechanisms determined to be significant for some applications; effective conventional programs for managing aging; and effective unconventional programs for managing aging. Since the AMG on heat exchangers was published four years ago, a brief review has been conducted to identify emerging regulatory issues, if any. The results of this review and lessons learned from the collective set of AMGs are presented.

Clauss, J.M.

1998-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Rewriting the standard on the functional requirements for computers used in safety systems of nuclear power plants  

SciTech Connect (OSTI)

Because of the rapid development of state-of-the-art computer technology, a rewrite of ANSI/IEEE-ANS-7-4.3.2-1982, {open_quotes}Application Criteria for Digital Computer Systems of Nuclear Power Generating Stations,{close_quotes} was required. This article outlines the thrust of this rewrite, which is nearing the balloting process, and identifies standards and guidelines to be used in the development of a highly reliable to be used in the development of a highly reliable computer system. The rewrite activity has been in process approximately 2 yr and is a cooperative project of the American Nuclear Society (ANS) Nuclear Power Plant Standards Committee (NUPPSCO) and the Institute of Electrical and Electronics Engineers (IEEE) Nuclear Power Engineering Committee (NPEC). Because computer technology has progressed significantly since ANSI/IEEE-ANS-7-4.3.2-1982 was issued, the rewrite was a very interesting challenge to the work group. The primary difference between the 1982 version and the rewrite is that the 1982 version addressed the quality assurance aspects of the Quality criteria, which included the integration of hardware and software and subsequent verification and validation, whereas the rewrite, being a product of IEEE Std 603-1991, {open_quotes}IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations,{close_quotes} was written to establish the minimum requirements for computer systems (hardware, software, and interfaces) used in safety systems. This article presents an introduction to the scope of the rewrite, provides a brief comment on how the work group addressed the scope, and follows with details on how the work group addressed the scope and rewrite direction. 1 fig., 1 tab.

Matras, J.R. [Science Applications International Corp., Las Vegas, NV (United States)

1991-07-01T23:59:59.000Z

342

Mechanistic understanding of irradiation-induced corrosion of zirconium alloys in nuclear power plants: Stimuli, status, and outlook  

SciTech Connect (OSTI)

Failures in the basic materials used in nuclear power plants continue to be costly and insidious, despite increasing industry vigilance to catch failures before they degrade safety. For instance, the overall costs to the US industry from materials problems could amount to as much as $10 billion annually. Moreover, estimates indicate that the cost of a pipe failure in a nuclear plant is one hundred times greater than the cost of a similar failure in a coal-fired plant. There are important practical stimuli and much scope for further understanding of the effects of irradiation on Zr-alloys (and other materials used in nuclear installations) by careful experimentation. Moreover, these studies need to address the effect of irradiation on all components of heterogeneous systems: the metal, the oxide and the environment, and especially those processes recurring at the interphases between these components. The present paper is aimed at providing specialists with some systematic information on the subject and with important considerations on the key items for further experimentation.

Johnson, A.B. Jr.; Ishigure, K.; Nechaev, A.F.; Reznichenko, E.A.; Cox, B.; Lemaignan, C.; Petrik, N.G.

1990-05-01T23:59:59.000Z

343

Qualification issues associated with the use of advanced instrumentation and control systems hardware in nuclear power plants  

SciTech Connect (OSTI)

The instrumentation and control (I&C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and Tiber-optic transmission. Elements of these advances in I&C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying I&C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I&C for present-day plants was compared to that proposed for advanced light water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light water reactor. The template was then used to address reliability issues for microprocessor-based protection systems. Standards (or lack thereof) for the qualification of microprocessor-based safety I&C systems were also identified. This approach addresses in part issues raised in Nuclear Regulatory Commission policy document SECY-91-292. which recognizes that advanced I&C systems for the nuclear industry are ``being developed without consensus standards, as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.``

Korsah, K. [Oak Ridge National Lab., TN (United States); Antonescu, C. [Nuclear Regulatory Commission, Rockville, MD (United States). Office of Nuclear Regulatory Research

1993-10-01T23:59:59.000Z

344

Experiments to investigate direct containment heating phenomena with scaled models of the Calvert Cliffs Nuclear Power Plant  

SciTech Connect (OSTI)

The Surtsey Test Facility is used to perform scaled experiments simulating High Pressure Melt Ejection accidents in a nuclear power plant (NPP). The experiments investigate the effects of direct containment heating (DCH) on the containment load. The results from Zion and Surry experiments can be extrapolated to other Westinghouse plants, but predicted containment loads cannot be generalized to all Combustion Engineering (CE) plants. Five CE plants have melt dispersal flow paths which circumvent the main mitigation of containment compartmentalization in most Westinghouse PWRs. Calvert Cliff-like plant geometries and the impact of codispersed water were addressed as part of the DCH issue resolution. Integral effects tests were performed with a scale model of the Calvert Cliffs NPP inside the Surtsey test vessel. The experiments investigated the effects of codispersal of water, steam, and molten core stimulant materials on DCH loads under prototypic accident conditions and plant configurations. The results indicated that large amounts of coejected water reduced the DCH load by a small amount. Large amounts of debris were dispersed from the cavity to the upper dome (via the annular gap). 22 refs., 84 figs., 30 tabs.

Blanchat, T.K.; Pilch, M.M.; Allen, M.D.

1997-02-01T23:59:59.000Z

345

The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code  

SciTech Connect (OSTI)

This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm{sup 2}, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

Sabundjian, Gaiane; Andrade, Delvonei A.; Belchior, Antonio Jr.; Silva Rocha, Marcelo da; Conti, Thadeu N.; Torres, Walmir M.; Macedo, Luiz A.; Umbehaun, Pedro E.; Mesquita, Roberto N.; Masotti, Paulo H. F.; Souza Lima, Ana Cecilia de [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN - SP) Av. Professor Lineu Prestes, 2242 05508-000 Sao Paulo, SP (Brazil)

2013-05-06T23:59:59.000Z

346

Recommended values for the distribution coefficient (Kd) to be used in dose assessments for decommissioning the Zion Nuclear Power Plant  

SciTech Connect (OSTI)

ZionSolutions is in the process of decommissioning the Zion Nuclear Power Plant. The site contains two reactor Containment Buildings, a Fuel Building, an Auxiliary Building, and a Turbine Building that may be contaminated. The current decommissioning plan involves removing all above grade structures to a depth of 3 feet below grade. The remaining underground structures will be backfilled. The remaining underground structures will contain low amounts of residual licensed radioactive material. An important component of the decommissioning process is the demonstration that any remaining activity will not cause a hypothetical individual to receive a dose in excess of 25 mrem/y as specified in 10CFR20 SubpartE.

Sullivan T.

2014-06-09T23:59:59.000Z

347

Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980  

SciTech Connect (OSTI)

This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

McCormack, K.E.; Gallaher, R.B.

1982-03-01T23:59:59.000Z

348

Aerial Survey Results for 131I Deposition on the Ground after the Fukushima Daiichi Nuclear Power Plant Accident  

SciTech Connect (OSTI)

In March 2011 the second largest accidental release of radioactivity in history occurred at the Fukushima Daiichi nuclear power plant following a magnitude 9.0 earthquake and subsequent tsunami. Teams from the U.S. Department of Energy, National Nuclear Security Administration Office of Emergency Response performed aerial surveys to provide initial maps of the dispersal of radioactive material in Japan. The initial results from the surveys did not report the concentration of 131I. This work reports on analyses performed on the initial survey data by a joint Japan-US collaboration to determine 131I ground concentration. This information is potentially useful in reconstruction of the inhalation and external exposure doses from this short-lived radionuclide. The deposited concentration of 134Cs is also reported.

Torii, Tatsuo [JAEA; Sugita, Takeshi [JAEA; Okada, Colin E. [NSTec; Reed, Michael S. [NSTec; Blumenthal, Daniel J. [NNSA

2013-08-01T23:59:59.000Z

349

Summary and bibliography of safety-related events at pressurized-water nuclear power plants as reported in 1979  

SciTech Connect (OSTI)

This report summarizes the data contained in reports submitted by licensees to the US Nuclear Regulatory Commission concerning safety-related operational events that occurred at pressurized-water-reactor nuclear power plants in 1979. A bibliography containing 100-word abstracts of the event reports is included. The 2064 abstracts included in the bibliography describe incidents, failures, and design or construction deficiencies experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Keyword and permuted-title indexes are provided to facilitate location of the abstracts of interest. Tables summarizing the information contained in the bibliography are also presented and discussed. Information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures, deficiencies noted, and time of occurrence (i.e., during refueling, operation, testing, or construction). Some of the more interesting events that occurred during the year are reviewed in detail. 33 refs.

Scott, R.L.; Gallaher, R.B.

1981-07-01T23:59:59.000Z

350

Testing of a naturally aged nuclear power plant inverter and battery charger  

SciTech Connect (OSTI)

A naturally aged inverter and battery charger were obtained from the Shippingport facility. This equipment was manufactured in 1974, and was installed at Shippingport in 1975 as part of a major plant modification. Testing was performed on this equipment under the auspices of the NRC's Nuclear Plant Aging Research (NPAR) Program to evaluate the type and extent of degradation due to aging, and to determine the effectiveness of condition monitoring techniques which could be used to detect aging effects. Steady state testing was conducted over the equipment's entire operating range. Step load changes were also initiated in order to monitor the electrical response. During this testing, component temperatures were monitored and circuit waveforms analyzed. Results indicated that aging had not substantially affected equipment operation. On the other hand, when compared with original acceptance test data, the monitoring techniques employed were sensitive to changes in measurable component and equipment parameters indicating the viability of detecting degradation prior to catastrophic failure. 7 refs., 34 figs., 12 tabs.

Gunther, W.E.

1988-09-01T23:59:59.000Z

351

NUCLEAR POWER in CALIFORNIA  

E-Print Network [OSTI]

NUCLEAR POWER in CALIFORNIA: 2007 STATUS REPORT CALIFORNIA ENERGY COMMISSION October 2007 CEC-100, California Contract No. 700-05-002 Prepared For: California Energy Commission Barbara Byron, Senior Nuclear public workshops on nuclear power. The Integrated Energy Policy Report Committee, led by Commissioners

352

Optimization and Comparison of Direct and Indirect Supercritical Carbon Dioxide Power Plant Cycles for Nuclear Applications  

SciTech Connect (OSTI)

There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550 C and 750 C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550 C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can be used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton cycle is the lower required operating temperature; 550 C versus 850 C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of both a direct and indirect supercritical CO2 Brayton Recompression cycle for different reactor outlet temperatures. The direct supercritical CO2 cycle transferred heat directly from a 600 MWt reactor to the supercritical CO2 working fluid supplied to the turbine generator at approximately 20 MPa. The indirect supercritical CO2 cycle assumed a helium-cooled Very High Temperature Reactor (VHTR), operating at a primary system pressure of approximately 7.0 MPa, delivered heat through an intermediate heat exchanger to the secondary indirect supercritical CO2 Brayton Recompression cycle, again operating at a pressure of about 20 MPa. For both the direct and indirect cycles, sensitivity calculations were performed for reactor outlet temperature between 550 C and 850 C. The UniSim models used realistic component parameters and operating conditions to model the complete reactor and power conversion systems. CO2 properties were evaluated, and the operating ranges of the cycles were adjusted to take advantage of the rapidly changing properties of CO2 near the critical point. The results of the analyses showed that, for the direct supercritical CO2 power cycle, thermal efficiencies in the range of 40 to 50% can be achieved. For the indirect supercritical CO2 power cycle, thermal efficiencies were approximately 10% lower than those obtained for the direct cycle over the same reactor outlet temperature range.

Edwin A. Harvego; Michael G. McKellar

2011-11-01T23:59:59.000Z

353

Features of adsorbed radioactive chemical elements and their isotopes distribution in iodine air filters AU-1500 at nuclear power plants  

E-Print Network [OSTI]

The main aim of research is to investigate the physical features of spatial distribution of the adsorbed radioactive chemical elements and their isotopes in the granular filtering medium in the iodine air filters of the type of AU1500 in the forced exhaust ventilation systems at the nuclear power plant. The gamma activation analysis method is applied to accurately characterize the distribution of the adsorbed radioactive chemical elements and their isotopes in the granular filtering medium in the AU1500 iodine air filter after its long term operation at the nuclear power plant. The typical spectrum of the detected chemical elements and their isotopes in the AU1500 iodine air filter, which was exposed to the bremsstrahlung gamma quantum irradiation, produced by the accelerating electrons in the tantalum target, are obtained. The spatial distributions of the detected chemical element 127I and some other chemical elements and their isotopes in the layer of absorber, which was made of the cylindrical coal granules of the type of SKT3, in the AU1500 iodine air filter are also researched. The possible influences by the standing acoustic wave of air pressure in the iodine air filter on the spatial distribution of the chemical elements and their isotopes in the iodine air filter are discussed. The comprehensive analysis of obtained research results on the distribution of the adsorbed chemical elements and their isotopes in the absorber of iodine air filter is performed.

I. M. Neklyudov; A. N. Dovbnya; N. P. Dikiy; O. P. Ledenyov; Yu. V. Lyashko

2013-06-21T23:59:59.000Z

354

Applying Human-performance Models to Designing and Evaluating Nuclear Power Plants: Review Guidance and Technical Basis  

SciTech Connect (OSTI)

Human performance models (HPMs) are simulations of human behavior with which we can predict human performance. Designers use them to support their human factors engineering (HFE) programs for a wide range of complex systems, including commercial nuclear power plants. Applicants to U.S. Nuclear Regulatory Commission (NRC) can use HPMs for design certifications, operating licenses, and license amendments. In the context of nuclear-plant safety, it is important to assure that HPMs are verified and validated, and their usage is consistent with their intended purpose. Using HPMs improperly may generate misleading or incorrect information, entailing safety concerns. The objective of this research was to develop guidance to support the NRC staff's reviews of an applicant's use of HPMs in an HFE program. The guidance is divided into three topical areas: (1) HPM Verification, (2) HPM Validation, and (3) User Interface Verification. Following this guidance will help ensure the benefits of HPMs are achieved in a technically sound, defensible manner. During the course of developing this guidance, I identified several issues that could not be addressed; they also are discussed.

O'Hara, J.M.

2009-11-30T23:59:59.000Z

355

Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants  

SciTech Connect (OSTI)

Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures.

Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

1995-05-01T23:59:59.000Z

356

Dose commitments due to radioactive releases from nuclear power plant sites in 1989  

SciTech Connect (OSTI)

Population and individual radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1989. Fifty-year dose commitments for a one-year exposure from both liquid and atmospheric releases were calculated for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each of 72 reactor sites. This report tabulates the results of these calculations, showing the dose commitments for both water and airborne pathways for each age group and organ. Also included for each of the sites is an estimate of individual doses which are compared with 10 CFR Part 50, Appendix I design objectives. The total collective dose commitments (from both liquid and airborne pathways) for each site ranged from a high of 14 person-rem to a low of 0.005 person-rem for the sites with plants in operation and producing power during the year. The arithmetic mean was 1.2 person-rem. The total population dose for all sites was estimated at 84 person-rem for the 140 million people considered at risk. The individual dose commitments estimated for all sites were below the Appendix I design objectives.

Baker, D.A. (Pacific Northwest Lab., Richland, WA (United States))

1993-02-01T23:59:59.000Z

357

RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

from the Rancho Seco nuclear plant was simulated, A total ofdistributions around the nuclear plant sites based on thegrowth surrounding nuclear plants after the issuance of the

Yen, W.W.S.

2010-01-01T23:59:59.000Z

358

CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

surrounding a nuclear plant, and they are stronglylocation for a nuclear plant, but it is the measures thatand consequences of nuclear plant accidents and would match

Nero, jA.V.

2010-01-01T23:59:59.000Z

359

The Fourth Generation of Nuclear Power  

SciTech Connect (OSTI)

The outlook for nuclear power in the U.S. is currently very bright. The economics, operations and safety performance of U.S. nuclear power plants is excellent. In addition, both the safety and economic regulation of nuclear power are being changed to produce better economic parameters for future nuclear plant operations and the licenses for plant operations are being extended to 60 years. There is further a growing awareness of the value of clean, emissions-free nuclear power. These parameters combine to form a firm foundation for continued successful U.S. nuclear plant operations, and even the potential In order to realize a bright future for nuclear power, we must respond successfully to five challenges: • Nuclear power must remain economically competitive, • The public must remain confident in the safety of the plants and the fuel cycle. • Nuclear wastes and spent fuel must be managed and the ultimate disposition pathways for nuclear wastes must be politically settled. • The proliferation potential of the commercial nuclear fuel cycle must continue to be minimized, and • We must assure a sustained manpower supply for the future and preserve the critical nuclear technology infrastructure. The Generation IV program is conceived to focus the efforts of the international nuclear community on responding to these challenges.

Lake, James Alan

2000-11-01T23:59:59.000Z

360

'Long-Cell Action' Corrosion: A Basic Mechanism Hidden Behind Components Degradation Issues in Nuclear Power Plants  

SciTech Connect (OSTI)

In spite of industries' effort over the last 40 years, corrosion-related issues continue to be one of the largest unresolved problems for nuclear power plants worldwide. There are several types of strange corrosion phenomena from the point of view of our current understanding of corrosion science established in other fields. Some of these are IGSCC, PWSCC, AOA, and FAC (Erosion-Corrosion). Through studying and coping with diverse corrosion phenomena, the author believes that they share a common basis with respect to the assumed corrosion mechanism (e.g., 'local cell action' hypothesis). In general, local cell action is rarely severe since it produces a fairly uniform corrosion. The 'long cell action' that transports electrons through structures far beyond the region of local cell corrosion activities has been identified as a basic mechanism in soil corrosion. If this mechanism is assumed in nuclear power plants, the structure becomes anodic in the area where the potential is less positive and cathodic where this potential is more positive. Metallic ions generated at anodic corrosion sites are transported to remote cathodic sites through the circulation of water and deposits as corrosion products. The SCC, FAC (E-C) and PWSCC occur in the anodic sites as the structure itself acts as a short-circuiting conductor between the two sites, the action is similar to a galvanic cell but in a very large scale. This situation is the same as a battery that has been short-circuited at the terminals. No apparent external potential difference exists between the two electrodes, but an electrochemical reaction is still taking place inside the battery cell with a large internal short current. In this example what is important is the potential difference between the local coolant and the surface of the structural material. Long cell action corrosion is likely enhancing the local cell action's anodic corrosion activities, such as SCC, FAC/E-C, and PWSCC. It tends to be more hazardous because of its localized nature compared with the local cell action corrosion. There exist various mechanisms (electrochemical cell configurations) that induce such potential differences, including: ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. In this paper, the author will discuss these potential differences and their relevance to the un-resolved corrosion issues in nuclear power plants. Due to the importance of this potential mechanism the author is calling for further verification experiments as a joint international project. (author)

Genn Saji [Ex-Secretariate of Nuclear Safety Commission of Japan (Japan)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology  

SciTech Connect (OSTI)

The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.

Kohut, P.; Epel, L.G.; Tutu, N.K. [and others

1998-08-01T23:59:59.000Z

362

Advanced Pipe Replacement Procedure for a Defective CRDM Housing Nozzle Enables Continued Normal Operation of a Nuclear Power Plant  

SciTech Connect (OSTI)

During the 2003 outage at the Ringhals Nuclear Plant in Sweden, a leak was found in the vicinity of a Control Rod Drive Mechanism (CRDM) housing nozzle at Unit 1. Based on the ALARA principle for radioactive contamination, a unique repair process was developed. The repair system includes utilization of custom, remotely controlled GTAW-robots, a CNC cutting and finishing machine, snake-arm robots and NDE equipment. The success of the repair solution was based on performing the machining and welding operations from the inside of the SCRAM pipe through the CRDM housing since accessibility from the outside was extremely limited. Before the actual pipe replacement procedure was performed, comprehensive training programs were conducted. Training was followed by certification of equipment, staff and procedures during qualification tests in a full scale mock-up of the housing nozzle. Due to the ingenuity of the overall repair solution and training programs, the actual pipe replacement procedure was completed in less than half the anticipated time. As a result of the successful pipe replacement, the nuclear power plant was returned to normal operation. (authors)

Gilmore, Geoff; Becker, Andrew [Climax Portable Machine Tools, Inc., 2712 East Second Street, Newberg, OR 97132 (United States)

2006-07-01T23:59:59.000Z

363

Developing Effective Continuous On-Line Monitoring Technologies to Manage Service Degradation of Nuclear Power Plants  

SciTech Connect (OSTI)

Recently, there has been increased interest in using prognostics (i.e, remaining useful life (RUL) prediction) for managing and mitigating aging effects in service-degraded passive nuclear power reactor components. A vital part of this philosophy is the development of tools for detecting and monitoring service-induced degradation. Experience with in-service degradation has shown that rapidly-growing cracks, including several varieties of stress corrosion cracks (SCCs), can grow through a pipe in less than one fuel outage cycle after they initiate. Periodic inspection has limited effectiveness at detecting and managing such degradation requiring a more versatile monitoring philosophy. Acoustic emission testing (AET) and guided wave ultrasonic testing (GUT) are related technologies with potential for on-line monitoring applications. However, harsh operating conditions within NPPs inhibit the widespread implementation of both technologies. For AET, another hurdle is the attenuation of passive degradation signals as they travel though large components, relegating AET to targeted applications. GUT is further hindered by the complexity of GUT signatures limiting its application to the inspection of simple components. The development of sensors that are robust and inexpensive is key to expanding the use of AET and GUT for degradation monitoring in NPPs and improving overall effectiveness. Meanwhile, the effectiveness of AET and GUT in NPPs can be enhanced through thoughtful application of tandem AET-GUT techniques.

Meyer, Ryan M.; Ramuhalli, Pradeep; Bond, Leonard J.; Cumblidge, Stephen E.

2011-09-30T23:59:59.000Z

364

Accident source terms for Light-Water Nuclear Power Plants. Final report  

SciTech Connect (OSTI)

In 1962 tile US Atomic Energy Commission published TID-14844, ``Calculation of Distance Factors for Power and Test Reactors`` which specified a release of fission products from the core to the reactor containment for a postulated accident involving ``substantial meltdown of the core``. This ``source term``, tile basis for tile NRC`s Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC`s reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ``source term`` release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ``source term`` is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it.

Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

1995-02-01T23:59:59.000Z

365

Occupational dose reduction at nuclear power plants: Annotated bibliography of selected readings in radiation protection and ALARA. Volume 7  

SciTech Connect (OSTI)

The ALARA Center at Brookhaven National Laboratory publishes a series of bibliographies of selected readings in radiation protection and ALARA in the continuing effort to collect and disseminate information on radiation dose reduction at nuclear power plants. This is volume 7 of the series. The abstracts in this bibliography were selected from proceedings of technical meetings and conferences, journals, research reports, and searches of the Energy Science and Technology database of the US Department of Energy. The subject material of these abstracts relates to radiation protection and dose reduction, and ranges from use of robotics to operational health physics, to water chemistry. Material on the design, planning, and management of nuclear power stations is included, as well as information on decommissioning and safe storage efforts. Volume 7 contains 293 abstract, an author index, and a subject index. The author index is specific for this volume. The subject index is cumulative and lists all abstract numbers from volumes 1 to 7. The numbers in boldface indicate the abstracts in this volume; the numbers not in boldface represent abstracts in previous volumes.

Kaurin, D.G.; Khan, T.A.; Sullivan, S.G.; Baum, J.W. [Brookhaven National Lab., Upton, NY (United States)

1993-07-01T23:59:59.000Z

366

Occupational dose reduction at nuclear power plants: Annotated bibliography of selected readings in radiation protection and ALARA. Volume 8  

SciTech Connect (OSTI)

The ALARA Center at Brookhaven National Laboratory publishes a series of bibliographies of selected readings in radiation protection and ALARA in a continuing effort to collect and disseminate information on radiation dose reduction at nuclear power plants. This volume 8 of the series. The abstracts in this bibliography were selected form proceedings of technical meetings and conference journals, research reports, and searches of the Energy Science and Technology database of the US Department of Energy. The subject material of these abstracts relates to the many aspects of radiation protection and dose reduction, and ranges form use of robotics, to operational health physics, to water chemistry. Material on the design, planning, and management of nuclear power stations is included, as well as information on decommissioning and safe storage efforts. Volume 8 contains 232 abstracts, an author index, and a subject index. The author index is specific for this volume. The subject index is cumulative and lists all abstract numbers from volumes 1 to 8. The numbers in boldface indicate the abstracts in this volume; the numbers not in boldface represent abstracts in previous volumes.

Sullivan, S.G.; Khan, T.A.; Xie, J.W. [Brookhaven National Lab., Upton, NY (United States)

1995-05-01T23:59:59.000Z

367

atr-fugen nuclear power: Topics by E-print Network  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Denmark December 1991 12;Abstract. A computer model of a simplified pressurized nuclear power plant a compute simulation of a simplified pressurized nuclear power plant model...

368

ASSESSMENT OF THE RADIONUCLIDE COMPOSITION OF "HOT PARTICLES" SAMPLED IN THE CHERNOBYL NUCLEAR POWER PLANT FOURTH REACTOR UNIT  

SciTech Connect (OSTI)

Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) 4th Reactor Unit Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified and the fuel burnup in these samples was determined. A systematic deviation in the burnup values based on the cesium isotopes, in comparison with other radionuclides, was observed. The conducted studies were the first ever performed to demonstrate the presence of significant quantities of {sup 242}Cm and {sup 243}Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from {sup 241}Am (and going higher), in comparison with the theoretical calculations.

Farfan, E.; Jannik, T.; Marra, J.

2011-10-01T23:59:59.000Z

369

Review of Recent Aging-Related Degradation Occurrences of Structures and Passive Components in U.S. Nuclear Power Plants  

SciTech Connect (OSTI)

The Korea Atomic Energy Research Institute (KAERI) and Brookhaven National Laboratory (BNL) are collaborating to develop seismic capability evaluation technology for degraded structures and passive components (SPCs) under a multi-year research agreement. To better understand the status and characteristics of degradation of SPCs in nuclear power plants (NPPs), the first step in this multi-year research effort was to identify and evaluate degradation occurrences of SPCs in U.S. NPPs. This was performed by reviewing recent publicly available information sources to identify and evaluate the characteristics of degradation occurrences and then comparing the information to the observations in the past. Ten categories of SPCs that are applicable to Korean NPPs were identified, comprising of anchorage, concrete, containment, exchanger, filter, piping system, reactor pressure vessel, structural steel, tank, and vessel. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

Nie,J.; Braverman, J.; Hofmayer, C.; Choun, Y.-S.; Kim, M.K.; Choi, I.-K.

2009-04-02T23:59:59.000Z

370

Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches  

SciTech Connect (OSTI)

The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

Steven R. Sherman

2007-06-01T23:59:59.000Z

371

Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework  

SciTech Connect (OSTI)

In the field of nuclear power plant (NPP) safety modeling, the perception of the role of socio-cognitive engineering (SCE) is continuously increasing. Today, the focus is especially on the identification of human and organization decisional errors caused by operators and managers under high-risk conditions, as evident by analyzing reports on nuclear incidents occurred in the past. At present, the engineering and social safety requirements need to enlarge their domain of interest in such a way to include all possible losses generating events that could be the consequences of an abnormal state of a NPP. Socio-cognitive modeling of Integrated Nuclear Safety Management (INSM) using the TOGA meta-theory has been discussed during the ICCAP 2011 Conference. In this paper, more detailed aspects of the cognitive decision-making and its possible human errors and organizational vulnerability are presented. The formal TOGA-based network model for cognitive decision-making enables to indicate and analyze nodes and arcs in which plant operators and managers errors may appear. The TOGA's multi-level IPK (Information, Preferences, Knowledge) model of abstract intelligent agents (AIAs) is applied. In the NPP context, super-safety approach is also discussed, by taking under consideration unexpected events and managing them from a systemic perspective. As the nature of human errors depends on the specific properties of the decision-maker and the decisional context of operation, a classification of decision-making using IPK is suggested. Several types of initial situations of decision-making useful for the diagnosis of NPP operators and managers errors are considered. The developed models can be used as a basis for applications to NPP educational or engineering simulators to be used for training the NPP executive staff. (authors)

Cappelli, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Gadomski, A. M. [ECONA, Centro Interuniversitario Elaborazione Cognitiva Sistemi Naturali e Artificiali, via dei Marsi 47, Rome (Italy); Sepiellis, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Wronikowska, M. W. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Poznan School of Social Sciences (Poland)

2012-07-01T23:59:59.000Z

372

Management of Ultimate Risk of Nuclear Power Plants by Source Terms - Lessons Learned from the Chernobyl Accident  

SciTech Connect (OSTI)

The term 'ultimate risk' is used here to describe the probabilities and radiological consequences that should be incorporated in siting, containment design and accident management of nuclear power plants for hypothetical accidents. It is closely related with the source terms specified in siting criteria which assures an adequate separation of radioactive inventories of the plants from the public, in the event of a hypothetical and severe accident situation. The author would like to point out that current source terms which are based on the information from the Windscale accident (1957) through TID-14844 are very outdated and do not incorporate lessons learned from either the Three Miles Island (TMI, 1979) nor Chernobyl accident (1986), two of the most severe accidents ever experienced. As a result of the observations of benign radionuclides released at TMI, the technical community in the US felt that a more realistic evaluation of severe reactor accident source terms was necessary. In this background, the 'source term research project' was organized in 1984 to respond to these challenges. Unfortunately, soon after the time of the final report from this project was released, the Chernobyl accident occurred. Due to the enormous consequences induced by then accident, the one time optimistic perspectives in establishing a more realistic source term were completely shattered. The Chernobyl accident, with its human death toll and dispersion of a large part of the fission fragments inventories into the environment, created a significant degradation in the public's acceptance of nuclear energy throughout the world. In spite of this, nuclear communities have been prudent in responding to the public's anxiety towards the ultimate safety of nuclear plants, since there still remained many unknown points revolving around the mechanism of the Chernobyl accident. In order to resolve some of these mysteries, the author has performed a scoping study of the dispersion and deposition mechanisms of fuel particles and fission fragments during the initial phase of the Chernobyl accident. Through this study, it is now possible to generally reconstruct the radiological consequences by using a dispersion calculation technique, combined with the meteorological data at the time of the accident and land contamination densities of {sup 137}Cs measured and reported around the Chernobyl area. Although it is challenging to incorporate lessons learned from the Chernobyl accident into the source term issues, the author has already developed an example of safety goals by incorporating the radiological consequences of the accident. The example provides safety goals by specifying source term releases in a graded approach in combination with probabilities, i.e. risks. The author believes that the future source term specification should be directly linked with safety goals. (author)

Genn Saji [Ex-Secretariate of Nuclear Safety Commission of Japan (Japan)

2006-07-01T23:59:59.000Z

373

Assessment of RELAP5/MOD2 code using loss of offsite power transient data of KNU (Korea Nuclear Unit) No. 1 Plant  

SciTech Connect (OSTI)

This report presents a code assessment study based on a real plant transient that occurred on June 9, 1981 at the KNU {number sign}1 (Korea Nuclear Unit Number 1). KNU {number sign}1 is a two-loop Westinghouse PWR plant of 587 Mwe. The loss of offsite power transient occurred at the 77.5% reactor power with 0.5%/hr power ramp. The real plant data were collected from available on-line plant records and computer diagnostics. The transient was simulated by RELAP5/MOD2/36.05 and the results were compared with the plant data to assess the code weaknesses and strengths. Some nodalization studies were performed to contribute to developing a guideline for PWR nodalization for the transient analysis. 5 refs., 18 figs., 3 tabs.

Chung, Bud-Dong; Kim, Hho-Jung (Korea Advanced Energy Research Inst., Daeduk-Danji (Republic of Korea). Korea Nuclear Safety Center); Lee, Young-Jin (Seoul National Univ. (Republic of Korea))

1990-04-01T23:59:59.000Z

374

Commercial nuclear power 1990  

SciTech Connect (OSTI)

This report presents the status at the end of 1989 and the outlook for commercial nuclear capacity and generation for all countries in the world with free market economies (FME). The report provides documentation of the US nuclear capacity and generation projections through 2030. The long-term projections of US nuclear capacity and generation are provided to the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) for use in estimating nuclear waste fund revenues and to aid in planning the disposal of nuclear waste. These projections also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment, and are provided to the Organization for Economic Cooperation and Development. The foreign nuclear capacity projections are used by the DOE uranium enrichment program in assessing potential markets for future enrichment contracts. The two major sections of this report discuss US and foreign commercial nuclear power. The US section (Chapters 2 and 3) deals with (1) the status of nuclear power as of the end of 1989; (2) projections of nuclear capacity and generation at 5-year intervals from 1990 through 2030; and (3) a discussion of institutional and technical issues that affect nuclear power. The nuclear capacity projections are discussed in terms of two projection periods: the intermediate term through 2010 and the long term through 2030. A No New Orders case is presented for each of the projection periods, as well as Lower Reference and Upper Reference cases. 5 figs., 30 tabs.

Not Available

1990-09-28T23:59:59.000Z

375

Flow Accelerated Erosion-Corrosion (FAC) considerations for secondary side piping in the AP1000{sup R} nuclear power plant design  

SciTech Connect (OSTI)

The issue of Flow Accelerated Erosion-Corrosion (FAC) in power plant piping is a known phenomenon that has resulted in material replacements and plant accidents in operating power plants. Therefore, it is important for FAC resistance to be considered in the design of new nuclear power plants. This paper describes the design considerations related to FAC that were used to develop a safe and robust AP1000{sup R} plant secondary side piping design. The primary FAC influencing factors include: - Fluid Temperature - Pipe Geometry/layout - Fluid Chemistry - Fluid Velocity - Pipe Material Composition - Moisture Content (in steam lines) Due to the unknowns related to the relative impact of the influencing factors and the complexities of the interactions between these factors, it is difficult to accurately predict the expected wear rate in a given piping segment in a new plant. This paper provides: - a description of FAC and the factors that influence the FAC degradation rate, - an assessment of the level of FAC resistance of AP1000{sup R} secondary side system piping, - an explanation of options to increase FAC resistance and associated benefits/cost, - discussion of development of a tool for predicting FAC degradation rate in new nuclear power plants. (authors)

Vanderhoff, J. F.; Rao, G. V. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Stein, A. [Shaw Power Nuclear, 1000 Technology Center Drive, Stoughton, MA 02072 (United States)

2012-07-01T23:59:59.000Z

376

Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1979  

SciTech Connect (OSTI)

This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1979. The 1345 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Keyword and permuted-title indexes to facilitate location of individual abstracts are provided in full size following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of the failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Some of the more interesting events that occurred during the year are reviewed in detail. 32 refs.

Scott, R.L.; Gallaher, R.B.

1981-06-01T23:59:59.000Z

377

System dynamics modeling for human performance in nuclear power plant operation  

E-Print Network [OSTI]

Perfect plant operation with high safety and economic performance is based on both good physical design and successful organization. However, in comparison with the affection that has been paid to technology research, the ...

Chu, Xinyuan

2006-01-01T23:59:59.000Z

378

A system-of-systems framework of Nuclear Power Plant Probabilistic Seismic Hazard Analysis by Fault Tree analysis and Monte Carlo simulation  

E-Print Network [OSTI]

A system-of-systems framework of Nuclear Power Plant Probabilistic Seismic Hazard Analysis by Fault and the Energetic Challenge, European Foundation for New Energy - Electricité de France, at �cole Centrale Paris - Supelec, France enrico.zio@ecp.fr, enrico.zio@supelec.fr b Department of Energy, Politecnico di Milano

Boyer, Edmond

379

Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown  

SciTech Connect (OSTI)

The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

Bylkin, Boris K. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Davydova, Galina B. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Zverkov, Yuri A. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Krayushkin, Alexander V. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Neretin, Yuri A. [Chernobyl Nuclear Power Plant (Ukraine); Nosovsky, Anatoly V. [Slavutych Division of the International Chernobyl Center (Ukraine); Seyda, Valery A. [Chernobyl Nuclear Power Plant (Ukraine); Short, Steven M. [Pacific Northwest National Laboratory (United States)

2001-10-15T23:59:59.000Z

380

Nuclear Power Generating Facilities (Maine)  

Broader source: Energy.gov [DOE]

The first subchapter of the statute concerning Nuclear Power Generating Facilities provides for direct citizen participation in the decision to construct any nuclear power generating facility in...

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings  

SciTech Connect (OSTI)

The Electricite de France`s N4 CHOOZ B nuclear power plant, two units of the world`s largest PWR model (1450 Mwe each), has earned the Electric Power International`s 1997 Powerplant Award. This lead NPP for EDF`s N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building`s inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter.

Rouelle, P.; Roy, F. [Electricite de France, Paris (France). Engineering and Construction Div.

1998-12-31T23:59:59.000Z

382

ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets  

E-Print Network [OSTI]

The affordable, robust, compact (ARC) reactor conceptual design study aims to reduce the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Q_p~13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ~63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ~23 T peak field on coil with newly available REBCO superconductor technology. External cu...

Sorbom, B N; Palmer, T R; Mangiarotti, F J; Sierchio, J M; Bonoli, P; Kasten, C; Sutherland, D A; Barnard, H S; Haakonsen, C B; Goh, J; Sung, C; Whyte, D G

2014-01-01T23:59:59.000Z

383

Space nuclear power and man's extraterrestrial civilization  

SciTech Connect (OSTI)

This paper examines leading space nuclear power technology candidates. Particular emphasis is given the heat-pipe reactor technology currently under development at the Los Alamos National Laboratory. This program is aimed at developing a 10-100 kWe, 7-year lifetime space nuclear power plant. As the demand for space-based power reaches megawatt levels, other nuclear reactor designs including: solid core, fluidized bed, and gaseous core, are considered.

Angelo, J.J.; Buden, D.

1983-01-01T23:59:59.000Z

384

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1  

SciTech Connect (OSTI)

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

NONE

1995-08-01T23:59:59.000Z

385

Feasibility Study of Hydrogen Production from Existing Nuclear Power Plants Using Alkaline Electrolysis  

SciTech Connect (OSTI)

The mid-range industrial market currently consumes 4.2 million metric tons of hydrogen per year and has an annual growth rate of 15% industries in this range require between 100 and 1000 kilograms of hydrogen per day and comprise a wide range of operations such as food hydrogenation, electronic chip fabrication, metals processing and nuclear reactor chemistry modulation.

Dana R. Swalla

2008-12-31T23:59:59.000Z

386

Using risk-based regulations for licensing nuclear power plants : case study of gas-cooled fast reactor  

E-Print Network [OSTI]

The strategy adopted for national energy supply is one of the most important policy choice for the US. Although it has been dismissed in the past decades, nuclear power today has key assets when facing concerns on energy ...

Jourdan, Grégoire

2005-01-01T23:59:59.000Z

387

Radioactive materials released from nuclear power plants: Annual report, 1993. Volume 14  

SciTech Connect (OSTI)

Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1993 have been compiled and reported. The summary data for the years 1974 through 1992 are included for comparison. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1993 release data are summarized in tabular form. Data covering specific radionuclides are summarized.

Tichler, J.; Doty, K.; Lucadamo, K. [Brookhaven National Lab., Upton, NY (United States)

1995-12-01T23:59:59.000Z

388

Update on the Cost of Nuclear Power  

E-Print Network [OSTI]

We update the cost of nuclear power as calculated in the MIT (2003) Future of Nuclear Power study. Our main focus is on the changing cost of construction of new plants. The MIT (2003) study provided useful data on the cost ...

Parsons, John E.

2009-01-01T23:59:59.000Z

389

Application of EPRI risk-based inservice inspection procedure to combustion engineering design of nuclear power plant  

SciTech Connect (OSTI)

The EPRI developed risk-based inservice inspection procedure is used to select the elements for inservice inspection on a section of the high pressure safety injection system of the Entergy Operations ANO2 nuclear plant. This plant is the pilot plant in a six utility-eleven plant EPRI tailored collaboration program to apply the general EPRI procedures to Combustion Engineering NSSS designs. The procedure results in a reduction of candidate inspection locations from 14, based on current ASME Section XI rules for B-J welds to 3, based on the risk-based selection criteria.

Lubin, B.T. [ABB Combustion Engineering, Windsor, CT (United States). Nuclear Operations; Fourgerousse, R. [Entergy Operations-ANO2, Russellville, AR (United States)

1996-12-01T23:59:59.000Z

390

Interim reliability evaluation program: analysis of the Arkansas Nuclear One. Unit 1 nuclear power plant. Volume 2 of 2  

SciTech Connect (OSTI)

Appendices are presented concerning systemic event tree analysis; analysis of the ANO-1 front-line and support systems; high pressure injection/high pressure recirculation system; low pressure injection/low pressure recirculation system; core flood system; reactor building spray system; emergency feedwater system; reactor building cooling system; reactor protection system; power conversion system; engineered safeguards activation system; service water system; class 1E AC power system; 125 volt DC system; battery and switchgear emergency cooling system; emergency feedwater initiation and control system; human interface system; sequence quantification; and supporting calculations.

Kolb, G.J.

1982-06-01T23:59:59.000Z

391

Protection against malevolent use of vehicles at Nuclear Power Plants. Vehicle barrier system selection guidance  

SciTech Connect (OSTI)

This manual provides a simplified procedure for selecting land vehicle barriers that will stop the design basis vehicle threat adopted by the U.S. Nuclear Regulatory Commission. Proper selection and construction of vehicle barriers should prevent intrusion of the design basis vehicle. In addition, vital safety related equipment should survive a design basis vehicle bomb attack when vehicle barriers are properly selected, sited, and constructed. This manual addresses passive vehicle barriers, active vehicle barriers, and site design features that can be used to reduce vehicle impact velocity.

Nebuda, D.T.

1994-08-01T23:59:59.000Z

392

Psychological scaling of expert estimates of human error probabilities: application to nuclear power plant operation  

SciTech Connect (OSTI)

The US Nuclear Regulatory Commission and Sandia National Laboratories sponsored a project to evaluate psychological scaling techniques for use in generating estimates of human error probabilities. The project evaluated two techniques: direct numerical estimation and paired comparisons. Expert estimates were found to be consistent across and within judges. Convergent validity was good, in comparison to estimates in a handbook of human reliability. Predictive validity could not be established because of the lack of actual relative frequencies of error (which will be a difficulty inherent in validation of any procedure used to estimate HEPs). Application of expert estimates in probabilistic risk assessment and in human factors is discussed.

Comer, K.; Gaddy, C.D.; Seaver, D.A.; Stillwell, W.G.

1985-01-01T23:59:59.000Z

393

Secretary Chu's Remarks at Vogtle Nuclear Power Plant -- As Prepared for  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administrationcontroller systemsBi (2) Sr (2)ScienceScientistsON THE5,to Visit Pantex

394

Ukraine Loads U.S. Nuclear Fuel into Power Plant as Part of DOE-Ukraine  

Broader source: Energy.gov (indexed) [DOE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists' Research Petroleum ReserveDepartment of EnergyDepartment ofRemarksNuclear Fuel

395

Advanced nuclear plant control complex  

DOE Patents [OSTI]

An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

1993-01-01T23:59:59.000Z

396

Wisconsin Nuclear Profile - Point Beach Nuclear Plant  

U.S. Energy Information Administration (EIA) Indexed Site

Point Beach Nuclear Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration...

397

Tennessee Nuclear Profile - Watts Bar Nuclear Plant  

U.S. Energy Information Administration (EIA) Indexed Site

Watts Bar Nuclear Plant" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration...

398

Structural aging program to assess the adequacy of critical concrete components in nuclear power plants  

SciTech Connect (OSTI)

The Structural Aging (SAG) Program is carried out by the Oak Ridge National Laboratory (ORNL) under sponsorship of the United States Nuclear Regulatory Commission (USNRC). The Program has evolved from preliminary studies conducted to evaluate the long-term environmental challenges to light-water reactor safety-related concrete civil structures. An important conclusion of these studies was that a damage methodology, which can provide a quantitative measure of a concrete structure's durability with respect to potential future requirements, needs to be developed. Under the SAG Program, this issue is being addressed through: establishment of a structural materials information center, evaluation of structural component assessment and repair technologies, and development of a quantitative methodology for structural aging determinations. Progress to date of each of these activities is presented as well as future plans. 7 refs., 5 figs.

Naus, D.J.; Marchbanks, M.F.; Oland, C.B.; Arndt, E.G.

1989-01-01T23:59:59.000Z

399

Computer–Based Procedures for Nuclear Power Plant Field Workers: Preliminary Results from Two Evaluation Studies  

SciTech Connect (OSTI)

The Idaho National Laboratory and participants from the U.S. nuclear industry are collaborating on a research effort aimed to augment the existing guidance on computer-based procedure (CBP) design with specific guidance on how to design CBP user interfaces such that they support procedure execution in ways that exceed the capabilities of paper-based procedures (PBPs) without introducing new errors. Researchers are employing an iterative process where the human factors issues and interface design principles related to CBP usage are systematically addressed and evaluated in realistic settings. This paper describes the process of developing a CBP prototype and the two studies conducted to evaluate the prototype. The results indicate that CBPs may improve performance by reducing errors, but may increase the time it takes to complete procedural tasks.

Katya L Le Blanc; Johanna H Oxstrand

2013-10-01T23:59:59.000Z

400

Using micro saint to predict performance in a nuclear power plant control room  

SciTech Connect (OSTI)

The United States Nuclear Regulatory Commission (NRC) requires a technical basis for regulatory actions. In the area of human factors, one possible technical basis is human performance modeling technology including task network modeling. This study assessed the feasibility and validity of task network modeling to predict the performance of control room crews. Task network models were built that matched the experimental conditions of a study on computerized procedures that was conducted at North Carolina State University. The data from the {open_quotes}paper procedures{close_quotes} conditions were used to calibrate the task network models. Then, the models were manipulated to reflect expected changes when computerized procedures were used. These models` predictions were then compared to the experimental data from the {open_quotes}computerized conditions{close_quotes} of the North Carolina State University study. Analyses indicated that the models predicted some subsets of the data well, but not all. Implications for the use of task network modeling are discussed.

Lawless, M.T.; Laughery, K.R. [Micro Analysis and Design, Inc., Boulder, CO (United States); Persenky, J.J. [Nuclear Regulatory Commission, Washington, DC (United States)

1995-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Nuclear power plant human computer interface design incorporating console simulation, operations personnel, and formal evaluation techniques  

SciTech Connect (OSTI)

New CRT-based information displays which enhance the human machine interface are playing a very important role and are being increasingly used in control rooms since they present a higher degree of flexibility compared to conventional hardwired instrumentation. To prototype a new console configuration and information display system at the Experimental Breeder Reactor II (EBR-II), an iterative process of console simulation and evaluation involving operations personnel is being pursued. Entire panels including selector switches and information displays are simulated and driven by plant dynamical simulations with realistic responses that reproduce the actual cognitive and physical environment. Careful analysis and formal evaluation of operator interaction while using the simulated console will be conducted to determine underlying principles for effective control console design for this particular group of operation personnel. Additional iterations of design, simulation, and evaluation will then be conducted as necessary.

Chavez, C.; Edwards, R.M.; Goldberg, J.H.

1993-12-31T23:59:59.000Z

402

Validation needs of seismic probabilistic risk assessment (PRA) methods applied to nuclear power plants  

SciTech Connect (OSTI)

An effort to validate seismic PRA methods is in progress. The work concentrates on the validation of plant response and fragility estimates through the use of test data and information from actual earthquake experience. Validation needs have been identified in the areas of soil-structure interaction, structural response and capacity, and equipment fragility. Of particular concern is the adequacy of linear methodology to predict nonlinear behavior. While many questions can be resolved through the judicious use of dynamic test data, other aspects can only be validated by means of input and response measurements during actual earthquakes. A number of past, ongoing, and planned testing programs which can provide useful validation data have been identified, and validation approaches for specific problems are being formulated.

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

1985-01-01T23:59:59.000Z

403

Decontamination of Radioactive Cesium Released from Fukushima Daiichi Nuclear Power Plant - 13277  

SciTech Connect (OSTI)

Peculiar binding of Cesium to the soil clay minerals remained the major obstacle for the immediate Cs-decontamination of soil and materials containing clay minerals like sludge. Experiments for the removal of Cesium from soil and ash samples from different materials were performed in the lab scale. For soil and sludge ash formed by the incineration of municipal sewage sludge, acid treatment at high temperature is effective while washing with water removed Cesium from ashes of plants or burnable garbage. Though total removal seems a difficult task, water-washing of wood-ash or garbage-ash at 40 deg. C removes >90% radiocesium, while >60% activity can be removed from soil and sludge-ash by acid washing at 95 deg. C. (authors)

Parajuli, Durga; Minami, Kimitaka; Tanaka, Hisashi; Kawamoto, Tohru [Nanosystem Research Institute, National Institute of Advanced Industrial Science and Technology - AIST (Japan)] [Nanosystem Research Institute, National Institute of Advanced Industrial Science and Technology - AIST (Japan)

2013-07-01T23:59:59.000Z

404

Removal of Radiocesium from Food by Processing: Data Collected after the Fukushima Daiichi Nuclear Power Plant Accident - 13167  

SciTech Connect (OSTI)

Removal of radiocesium from food by processing is of great concern following the accident of TEPCO's Fukushima Daiichi Nuclear Power Plant accident. Foods in markets are monitored and recent monitoring results have shown that almost all food materials were under the standard limit concentration levels for radiocesium (Cs-134+137), that is, 100 Bq kg{sup -1} in raw foods, 50 Bq kg{sup -1} in baby foods, and 10 Bq kg{sup -1} in drinking water; those food materials above the limit cannot be sold. However, one of the most frequently asked questions from the public is how much radiocesium in food would be removed by processing. Hence, information about radioactivity removal by processing of food crops native to Japan is actively sought by consumers. In this study, the food processing retention factor, F{sub r}, which is expressed as total activity in processed food divided by total activity in raw food, is reported for various types of corps. For white rice at a typical polishing yield of 90-92% from brown rice, the F{sub r} value range was 0.42-0.47. For leafy vegetable (indirect contamination), the average F{sub r} values were 0.92 (range: 0.27-1.2) after washing and 0.55 (range: 0.22-0.93) after washing and boiling. The data for some fruits are also reported. (authors)

Uchida, Shigeo; Tagami, Keiko [Office of Biospheric Assessment for Waste Disposal, National Institute of Radiological Sciences, Anagawa 4-9-1, Inage-ku, Chiba 263-8555 (Japan)] [Office of Biospheric Assessment for Waste Disposal, National Institute of Radiological Sciences, Anagawa 4-9-1, Inage-ku, Chiba 263-8555 (Japan)

2013-07-01T23:59:59.000Z

405

An overview of environmental degradation of materials in nuclear power plant piping systems  

SciTech Connect (OSTI)

Piping in light water reactor (LWR) power systems is affected by several types of environmental degradation: intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs) has required research, inspection, and mitigation programs that will ultimately cost several billion dollars; erosion-corrosion of carbon steel piping has been observed frequently in the secondary systems of both BWRs and pressurized water reactors (PWRs); the effect of the BWR environment can greatly diminish the design margin inherent in the ASME Section III fatigue design curves for carbon steel piping; and cast stainless steels are subject to embrittlement after extended thermal aging at reactor operating temperatures. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions.

Shack, W.J.

1987-08-01T23:59:59.000Z

406

Radiation Testing of a Low Voltage Silicone Nuclear Power Plant Cable.  

SciTech Connect (OSTI)

This report summarizes the results generated in FY13 for cable insulation in support of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program, in collaboration with the US-Argentine Binational Energy Working Group (BEWG). A silicone (SiR) cable, which was stored in benign conditions for ~30 years, was obtained from Comision Nacional de Energia Atomica (CNEA) in Argentina with the approval of NA-SA (Nucleoelectrica Argentina Sociedad Anonima). Physical property testing was performed on the as-received cable. This cable was artificially aged to assess behavior with additional analysis. SNL observed appreciable tensile elongation values for all cable insulations received, indicative of good mechanical performance. Of particular note, the work presented here provides correlations between measured tensile elongation and other physical properties that may be potentially leveraged as a form of condition monitoring (CM) for actual service cables. It is recognized at this point that the polymer aging community is still lacking the number and types of field returned materials that are desired, but Sandia National Laboratories (SNL) -- along with the help of others -- is continuing to work towards that goal. This work is an initial study that should be complimented with location-mapping of environmental conditions of Argentinean plant conditions (dose and temperature) as well as retrieval, analysis, and comparison with in- service cables.

White II, Gregory Von; Schroeder, John Lee.; Sawyer, Patricia Sue.; Wichhart, Derek; Mata, Guillermo Adrian; Zorrilla, Jorge; Bernstein, Robert

2014-09-01T23:59:59.000Z

407

Radiation Testing of a Low Voltage Silicone Nuclear Power Plant Cable.  

SciTech Connect (OSTI)

This report summarizes the results generated in FY13 for cable insulation in support of DOE's Light Water Reactor Sustainability (LWRS) Program, in collaboration with the US- Argentine Binational Energy Working Group (BEWG). A silicone (SiR) cable, which was stored in benign conditions for ~30 years, was obtained from Comision Nacional de Energia Atomica (CNEA) in Argentina. Physical property testing was performed on the as-received cable. This cable was artificially aged to assess behavior with additional analysis. SNL observed appreciable tensile elongation values for all cable insulations received, indicative of good mechanical performance. Of particular note, the work presented here provides correlations between measured tensile elongation and other physical properties that may be potentially leveraged as a form of condition monitoring (CM) for actual service cables. It is recognized at this point that the polymer aging community is still lacking the number and types of field returned materials that are desired, but SNL -- along with the help of others -- is continuing to work towards that goal. This work is an initial study that should be complimented with location- mapping of environmental conditions of CNEA plant conditions (dose and temperature) as well as retrieval, analysis, and comparison with in-service cables.

Bernstein, Robert

2014-08-01T23:59:59.000Z

408

Experimental results from pressure testing a 1:6-scale nuclear power plant containment  

SciTech Connect (OSTI)

This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission`s program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix.

Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

1992-01-01T23:59:59.000Z

409

Evaluation of the Effectiveness of a New Technology for Extraction of Insoluble Impurities from Nuclear Power Plant Steam Generators with Purge Water  

SciTech Connect (OSTI)

An experimental technology for the removal of insoluble impurities from a horizontal steam generator with purge water during planned shutdowns of the power generating unit is improved through a more representative determination of the concentration of impurities in the purge water ahead of the water cleanup facility and a more precise effective time for the duration of the purge process. Tests with the improved technique at power generating unit No. 1 of the Rostov Nuclear Power Plant show that the efficiency with which insoluble impurities are removed from the steam generator volume was more than two orders of magnitude greater than under the standard regulations.

Bud'ko, I. O. [JSC NIITsE 'Tsentrenergo' (Russian Federation)] [JSC NIITsE 'Tsentrenergo' (Russian Federation); Zhukov, A. G. [Rostov Nuclear Power Plant (Russian Federation)] [Rostov Nuclear Power Plant (Russian Federation)

2013-11-15T23:59:59.000Z

410

Geothermal Power Plants — Minimizing Land Use and Impact  

Broader source: Energy.gov [DOE]

For energy production and development, geothermal power plants don't use much land compared to coal and nuclear power plants. And the environmental impact upon the land they use is minimal.

411

Incentive Cost Recovery Rule for Nuclear Power Generation (Louisiana)  

Broader source: Energy.gov [DOE]

The Incentive Cost Recovery Rule for Nuclear Power Generation establishes guidelines for any utility seeking to develop a nuclear power plant in Louisiana. The rule clarifies, as well as...

412

Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.  

SciTech Connect (OSTI)

The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

2013-05-01T23:59:59.000Z

413

Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants – Interim Study FY13  

SciTech Connect (OSTI)

The most important criterion for cable performance is its ability to withstand a design-basis accident. With nearly 1000 km of power, control, instrumentation, and other cables typically found in an NPP, it would be a significant undertaking to inspect all of the cables. Degradation of the cable jacket, electrical insulation, and other cable components is a key issue that is likely to affect the ability of the currently installed cables to operate safely and reliably for another 20 to 40 years beyond the initial operating life. The development of one or more nondestructive evaluation (NDE) techniques and supporting models that could assist in determining the remaining life expectancy of cables or their current degradation state would be of significant interest. The ability to nondestructively determine material and electrical properties of cable jackets and insulation without disturbing the cables or connections has been deemed essential. Currently, the only technique accepted by industry to measure cable elasticity (the gold standard for determining cable insulation degradation) is the indentation measurement. All other NDE techniques are used to find flaws in the cable and do not provide information to determine the current health or life expectancy. There is no single NDE technique that can satisfy all of the requirements needed for making a life-expectancy determination, but a wide range of methods have been evaluated for use in NPPs as part of a continuous evaluation program. The commonly used methods are indentation and visual inspection, but these are only suitable for easily accessible cables. Several NDE methodologies using electrical techniques are in use today for flaw detection but there are none that can predict the life of a cable. There are, however, several physical and chemical ptoperty changes in cable insulation as a result of thermal and radiation damage. In principle, these properties may be targets for advanced NDE methods to provide early warning of aging and degradation. Examples of such key indicators include changes in chemical structure, mechanical modulus, and dielectric permittivity. While some of these indicators are the basis of currently used technologies, there is a need to increase the volume of cable that may be inspected with a single measurement, and if possible, to develop techniques for in-situ inspection (i.e., while the cable is in operation). This is the focus of the present report.

Simmons, Kevin L.; Fifield, Leonard S.; Westman, Matthew P.; Ramuhalli, Pradeep; Pardini, Allan F.; Tedeschi, Jonathan R.; Jones, Anthony M.

2013-09-27T23:59:59.000Z

414

GEOTHERMAL POWER GENERATION PLANT  

SciTech Connect (OSTI)

Oregon Institute of Technology (OIT) drilled a deep geothermal well on campus (to 5,300 feet deep) which produced 196oF resource as part of the 2008 OIT Congressionally Directed Project. OIT will construct a geothermal power plant (estimated at 1.75 MWe gross output). The plant would provide 50 to 75 percent of the electricity demand on campus. Technical support for construction and operations will be provided by OIT’s Geo-Heat Center. The power plant will be housed adjacent to the existing heat exchange building on the south east corner of campus near the existing geothermal production wells used for heating campus. Cooling water will be supplied from the nearby cold water wells to a cooling tower or air cooling may be used, depending upon the type of plant selected. Using the flow obtained from the deep well, not only can energy be generated from the power plant, but the “waste” water will also be used to supplement space heating on campus. A pipeline will be construction from the well to the heat exchanger building, and then a discharge line will be construction around the east and north side of campus for anticipated use of the “waste” water by facilities in an adjacent sustainable energy park. An injection well will need to be drilled to handle the flow, as the campus existing injection wells are limited in capacity.

Boyd, Tonya

2013-12-01T23:59:59.000Z

415

Nuclear power plant cable materials : review of qualification and currently available aging data for margin assessments in cable performance.  

SciTech Connect (OSTI)

A selective literature review was conducted to assess whether currently available accelerated aging and original qualification data could be used to establish operational margins for the continued use of cable insulation and jacketing materials in nuclear power plant environments. The materials are subject to chemical and physical degradation under extended radiationthermal- oxidative conditions. Of particular interest were the circumstances under which existing aging data could be used to predict whether aged materials should pass loss of coolant accident (LOCA) performance requirements. Original LOCA qualification testing usually involved accelerated aging simulations of the 40-year expected ambient aging conditions followed by a LOCA simulation. The accelerated aging simulations were conducted under rapid accelerated aging conditions that did not account for many of the known limitations in accelerated polymer aging and therefore did not correctly simulate actual aging conditions. These highly accelerated aging conditions resulted in insulation materials with mostlyinert' aging processes as well as jacket materials where oxidative damage dropped quickly away from the air-exposed outside jacket surface. Therefore, for most LOCA performance predictions, testing appears to have relied upon heterogeneous aging behavior with oxidation often limited to the exterior of the cable cross-section - a situation which is not comparable with the nearly homogenous oxidative aging that will occur over decades under low dose rate and low temperature plant conditions. The historical aging conditions are therefore insufficient to determine with reasonable confidence the remaining operational margins for these materials. This does not necessarily imply that the existing 40-year-old materials would fail if LOCA conditions occurred, but rather that unambiguous statements about the current aging state and anticipated LOCA performance cannot be provided based on original qualification testing data alone. The non-availability of conclusive predictions for the aging conditions of 40-year-old cables implies that the same levels of uncertainty will remain for any re-qualification or extended operation of these cables. The highly variable aging behavior of the range of materials employed also implies that simple, standardized aging tests are not sufficient to provide the required aging data and performance predictions for all materials. It is recommended that focused studies be conducted that would yield the material aging parameters needed to predict aging behaviors under low dose, low temperature plant equivalent conditions and that appropriately aged specimens be prepared that would mimic oxidatively-aged 40- to 60- year-old materials for confirmatory LOCA performance testing. This study concludes that it is not sufficient to expose materials to rapid, high radiation and high temperature levels with subsequent LOCA qualification testing in order to predictively quantify safety margins of existing infrastructure with regard to LOCA performance. We need to better understand how cable jacketing and insulation materials have degraded over decades of power plant operation and how this aging history relates to service life prediction and the performance of existing equipment to withstand a LOCA situation.

Celina, Mathias Christopher; Gillen, Kenneth Todd; Lindgren, Eric Richard

2013-05-01T23:59:59.000Z

416

Physical features of small disperse coal dust fraction transportation and structurization processes in iodine air filters of absorption type in ventilation systems at nuclear power plants  

E-Print Network [OSTI]

The research on the physical features of transportation and structurization processes by the air-dust aerosol in the granular filtering medium with the cylindrical coal adsorbent granules in an air filter of the adsorption type in the heating ventilation and cooling (HVAC) system at the nuclear power plant is completed. The physical origins of the coal dust masses distribution along the absorber with the granular filtering medium with the cylindrical coal granules during the air-dust aerosol intake process in the near the surface layer of absorber are researched. The quantitative technical characteristics of air filtering elements, which have to be considered during the optimization of air filters designs for the application in the ventilation systems at the nuclear power plants, are obtained.

Ledenyov, Oleg P; Poltinin, P Ya; Fedorova, L I

2012-01-01T23:59:59.000Z

417

Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models  

SciTech Connect (OSTI)

Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs.

Evans, J.S.; Moeller, D.W.; Cooper, D.W.

1985-07-01T23:59:59.000Z

418

LONG-TERM DYNAMICS OF RADIONUCLIDE VERTICAL MIGRATION IN SOILS OF THE CHERNOBYL NUCLEAR POWER PLANT EXCLUSION ZONE  

SciTech Connect (OSTI)

The radioactive fallout from the Chernobyl Nuclear Power Plant (ChNPP) accident consisted of fuel and condensation components. An important radioecological task associated with the late phase of the accident is to evaluate the dynamics of radionuclide mobility in soils. Identification of the variability (or invariability) in the radionuclide transfer parameters makes it possible to (1) accurately predict migration patterns and biological availability of radionuclides and (2) evaluate long-term exposure trends for the population who may reoccupy the remediated abandoned areas. In 1986-1987, a number of experimental plots were established within various tracts of the fallout plume to assist with the determination of the long-term dynamics of radionuclide vertical migration in the soils. The transfer parameters for {sup 137}Cs, {sup 90}Sr, and {sup 239,240}Pu in the soil profile, as well as their ecological half-time of the radionuclide residence (T{sub 1/2}{sup ecol}) values in the upper 5-cm thick soil layers of different grasslands were estimated at various times since the accident. Migration characteristics in the grassland soils tend to decrease as follows: {sup 90}Sr > {sup 137}Cs {ge} {sup 239,240}Pu. It was found that the {sup 137}Cs absolute T{sub 1/2}{sup ecol} values are 3-7 times higher than its radioactive decay half-life value. Therefore, changes in the exposure dose resulting from the soil deposited {sup 137}Cs now depend only on its radioactive decay. The {sup 90}Sr T{sub 1/2}{sup ecol} values for the 21st year after the fallout tend to decrease, indicating an intensification of its migration capabilities. This trend appears consistent with a pool of mobile {sup 90}Sr forms that grows over time due to destruction of the fuel particles.

Farfan, E.

2009-11-19T23:59:59.000Z

419

ENGINEERED NEAR SURFACE DISPOSAL FACILITY OF THE INDUSTRIAL COMPLEX FOR SOLID RADWASTE MANAGEMENT AT CHERNOBYL NUCLEAR POWER PLANT  

SciTech Connect (OSTI)

As a part of the turnkey project ''Industrial Complex for Solid Radwaste Management (ICSRM) at the Chernobyl Nuclear Power Plant (ChNPP)'' an Engineered Near Surface Disposal Facility (ENSDF, LOT 3) will be built on the VEKTOR site within the 30 km Exclusion Zone of the ChNPP. This will be performed by RWE NUKEM GmbH, Germany, and it governs the design, licensing support, fabrication, assembly, testing, inspection, delivery, erection, installation and commissioning of the ENSDF. The ENSDF will receive low to intermediate level, short lived, processed/conditioned wastes from the ICSRM Solid Waste Processing Facility (SWPF, LOT 2), the ChNPP Liquid Radwaste Treatment Plant (LRTP) and the ChNPP Interim Storage Facility for RBMK Fuel Assemblies (ISF). The ENSDF has a capacity of 55,000 m{sup 3}. The primary functions of the ENSDF are: to receive, monitor and record waste packages, to load the waste packages into concrete disposal units, to enable capping and closure of the disposal unit s, to allow monitoring following closure. The ENSDF comprises the turnkey installation of a near surface repository in the form of an engineered facility for the final disposal of LILW-SL conditioned in the ICSRM SWPF and other sources of Chernobyl waste. The project has to deal with the challenges of the Chernobyl environment, the fulfillment of both Western and Ukrainian standards, and the installation and coordination of an international project team. It will be shown that proven technologies and processes can be assembled into a unique Management Concept dealing with all the necessary demands and requirements of a turnkey project. The paper emphasizes the proposed concepts for the ENSDF and their integration into existing infrastructure and installations of the VEKTOR site. Further, the paper will consider the integration of Western and Ukrainian Organizations into a cohesive project team and the requirement to guarantee the fulfillment of both Western standards and Ukrainian regulations and licensing requirements. The paper provides information on the output of the Detail Design and will reflect the progress of the design work.

Ziehm, Ronny; Pichurin, Sergey Grigorevich

2003-02-27T23:59:59.000Z

420

Identification and Assessment of Material Models for Age-Related Degradation of Structures and Passive Components in Nuclear Power Plants  

SciTech Connect (OSTI)

When performing seismic safety assessments of nuclear power plants (NPPs), the potential effects of age-related degradation on structures, systems, and components (SSCs) should be considered. To address the issue of aging degradation, the Korea Atomic Energy Research Institute (KAERI) has embarked on a five-year research project to develop a realistic seismic risk evaluation system which will include the consideration of aging of structures and components in NPPs. Three specific areas that are included in the KAERI research project, related to seismic probabilistic risk assessment (PRA), are probabilistic seismic hazard analysis, seismic fragility analysis including the effects of aging, and a plant seismic risk analysis. To support the development of seismic capability evaluation technology for degraded structures and components, KAERI entered into a collaboration agreement with Brookhaven National Laboratory (BNL) in 2007. The collaborative research effort is intended to continue over a five year period with the goal of developing seismic fragility analysis methods that consider the potential effects of age-related degradation of SSCs, and using these results as input to seismic PRAs. In the Year 1 scope of work BNL collected and reviewed degradation occurrences in US NPPs and identified important aging characteristics needed for the seismic capability evaluations that will be performed in the subsequent evaluations in the years that follow. This information is presented in the Annual Report for the Year 1 Task, identified as BNL Report-81741-2008 and also designated as KAERI/RR-2931/2008. The report presents results of the statistical and trending analysis of this data and compares the results to prior aging studies. In addition, the report provides a description of U.S. current regulatory requirements, regulatory guidance documents, generic communications, industry standards and guidance, and past research related to aging degradation of SSCs. This report describes the research effort performed by BNL for the Year 2 scope of work. This research focused on methods that could be used to represent the long-term behavior of materials used at NPPs. To achieve this BNL reviewed time-dependent models which can approximate the degradation effects of the key materials used in the construction of structures and passive components determined to be of interest in the Year 1 effort. The intent was to review the degradation models that would cover the most common time-dependent changes in material properties for concrete and steel components.

Nie,J.; Braverman, J.; Hofmayer, C.; Kim, M. K.; Choi, I-K.

2009-04-27T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear power plant" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Issues for New Nuclear Plants  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

to produce heavy components and nuclear-grade equipment - Transportation of heavy components - Constructionoperation workforce - Cost of new plants Cooling Technology...

422

The American nuclear power industry. A handbook  

SciTech Connect (OSTI)

This book presents an overview of the history and current organization of the American nuclear power industry. Part I focuses on development of the industry, including the number, capacity, and type of plants in commercial operation as well as those under construction. Part II examines the safety, environmental, antitrust, and licensing issues involved in the use of nuclear power. Part III presents case studies of selected plants, such as Three Mile Island and Seabrook, to illustrate some of the issues discussed. The book also contains a listing of the Nuclear Regulatory Commission libraries and a subject index.

Pearman, W.A.; Starr, P.

1984-01-01T23:59:59.000Z

423

LIFE Power Plant Fusion Power Associates  

E-Print Network [OSTI]

LIFE Power Plant Fusion Power Associates December 14, 2011 Mike Dunne LLNL #12;NIf-1111-23714.ppt LIFE power plant 2 #12;LIFE delivery timescale NIf-1111-23714.ppt 3 #12;Timely delivery is enabled dpa) § Removes ion threat and mitigates x-ray threat ­ allows simple steel piping § No need

424

Overview paper on nuclear power  

SciTech Connect (OSTI)

This paper was prepared as an input to ORNL's Strategic Planning Activity, ORNL National Energy Perspective (ONEP). It is intended to provide historical background on nuclear power, an analysis of the mission of nuclear power, a discussion of the issues, the technology choices, and the suggestion of a strategy for encouraging further growth of nuclear power.

Spiewak, I.; Cope, D.F.

1980-09-01T23:59:59.000Z

425

Massachusetts Nuclear Profile - Pilgrim Nuclear Power Station  

U.S. Energy Information Administration (EIA) Indexed Site

Pilgrim Nuclear Power Station" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer cpacity factor (percent)","Type","Commercial operation date","License...

426

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons  

SciTech Connect (OSTI)

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

Ioffe, B. L.; Kochurov, B. P. [Institute of Theoretical and Experimental Physics (Russian Federation)

2012-02-15T23:59:59.000Z

427

Guidance for Deployment of Mobile Technologies for Nuclear Power...  

Broader source: Energy.gov (indexed) [DOE]

This report is a guidance document prepared for the benefit of commercial nuclear power plants' (NPPs) supporting organizations and personnel who are considering or undertaking...

428

Removal of Radionuclides from Waste Water at Fukushima Daiichi Nuclear Power Plant: Desalination and Adsorption Methods - 13126  

SciTech Connect (OSTI)

Waste water containing high levels of radionuclides due to the Fukushima Daiichi Nuclear Power Plant accident, has been treated by the adsorption removal and reverse-osmosis (RO) desalination to allow water re-use for cooling the reactors. Radionuclides in the waste water are collected in the adsorbent medium and the RO concentrate (RO brine) in the water treatment system currently operated at the Fukushima Daiichi site. In this paper, we have studied the behavior of radionuclides in the presently applied RO desalination system and the removal of radionuclides in possible additional adsorption systems for the Fukushima Daiichi waste water treatment. Regarding the RO desalination system, decontamination factors (DFs) of the elements present in the waste water were obtained by lab-scale testing using an RO unit and simulated waste water with non-radioactive elements. The results of the lab-scale testing using representative elements showed that the DF for each element depended on its hydrated ionic radius: the larger the hydrated ionic radius of the element, the higher its DF is. Thus, the DF of each element in the waste water could be estimated based on its hydrated ionic radius. For the adsorption system to remove radionuclides more effectively, we studied adsorption behavior of typical elements, such as radioactive cesium and strontium, by various kinds of adsorbents using batch and column testing. We used batch testing to measure distribution coefficients (K{sub d}s) for cesium and strontium onto adsorbents under different brine concentrations that simulated waste water conditions at the Fukushima Daiichi site. For cesium adsorbents, K{sub d}s with different dependency on the brine concentration were observed based on the mechanism of cesium adsorption. As for strontium, K{sub d}s decreased as the brine concentration increased for any adsorbents which adsorbed strontium by intercalation and by ion exchange. The adsorbent titanium oxide had higher K{sub d}s and it was used for the column testing to obtain breakthrough curves under various conditions of pH and brine concentration. The breakthrough point had a dependency on pH and the brine concentration. We found that when the pH was higher or the brine concentration was lower, the longer it took to reach the breakthrough point. The inhibition of strontium adsorption by alkali earth metals would be diminished for conditions of higher pH and lower brine concentration. (authors)

Kani, Yuko; Kamosida, Mamoru; Watanabe, Daisuke [Hitachi Research Laboratory, Hitachi, Ltd., 7-2-1 Omika-cho, Hitachi, Ibaraki, 319-1221 (Japan)] [Hitachi Research Laboratory, Hitachi, Ltd., 7-2-1 Omika-cho, Hitachi, Ibaraki, 319-1221 (Japan); Asano, Takashi; Tamata, Shin [Hitachi Works, Hitachi-GE Nuclear Energy, Ltd. (Japan)] [Hitachi Works, Hitachi-GE Nuclear Energy, Ltd. (Japan)

2013-07-01T23:59:59.000Z

429

POWER PLANT RELIABILITY-AVAILABILITY AND STATE REGULATION. VOLUME 7 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Removal Equipment (nuclear plant) Turbine Building ClosedCooling Water System (nuclear plant) SteamReheater (nuclear plant) Inspection Water Induction

Nero, A.V.

2010-01-01T23:59:59.000Z

430

Economics of nuclear power in Finland  

SciTech Connect (OSTI)

The nuclear power generation fits perfectly with the long duration load profile of the Finnish power system. The good performance of the Finnish nuclear power has yielded benefits also to the consumers through its contribution to decreasing the electricity price. Furthermore, the introduction of nuclear power has resulted in a clear drop in carbon dioxide emissions from electricity generation in the shift of 1970's and 1980's. In the year 2001 the four Finnish nuclear power units at Loviisa and Olkiluoto generated 22.8 TWh electricity, equivalent to 28 per cent of the total consumption. Loviisa power station has a net output capacity of 2 x 488 MW, and Olkiluoto 2 x 840 MW. The capacity factors of the four nuclear units have been above 90 per cent, which are among the highest in the world. The energy-intensive process industries in particular have strong belief in nuclear power. In November 2000, Teollisuuden Voima company (TVO) submitted to the Finnish Government an application for decision in principle concerning the construction of a new nuclear power plant unit. The arguments were among other things to guarantee for the Finnish industry the availability of cheap electric energy and to meet the future growth of electricity consumption in Finland. The carbon-free nuclear power also represents the most efficient means to meet the Greenhouse Gas abatement quota of Finland. Simultaneously, the energy policy of the Government includes intensive R and D and investment support for the renewable energy sources and energy conservation, and the objective is also to replace coal with natural gas as much as reasonably possible. The fifth nuclear unit would be located in one of the existing Finnish nuclear sites, i.e. Olkiluoto or Loviisa. The size of the new nuclear unit would be in the range of 1000 to 1600 MW electric. The ready infrastructure of the existing site could be utilised resulting in lower investment cost for the new unit. The Finnish Government accepted the application of TVO Company on January 17, 2002, but the final word will be said by the Parliament. During the spring 2002 there will be intensive discussion on all levels, whether nuclear power is for or against 'the total benefit of the society'. The Parliament decision is expected to be made by the summer 2002. In this paper, firstly a financial comparison of the new base-load power plant alternatives is carried out in the Finnish circumstances, and secondly the actual power production costs of the existing Olkiluoto nuclear power plant based on the operating history of about 20 years will be referred. (authors)

Tarjanne, Risto; Luostarinen, Kari [Lappeenranta University of Technology, Department of Energy and Environmental Technology, PO Box 20, FIN-53851 Lappeenranta (Finland)

2002-07-01T23:59:59.000Z

431

Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants  

SciTech Connect (OSTI)

The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

2012-09-14T23:59:59.000Z

432

The Fukushima Nuclear Event and its Implications for Nuclear Power  

SciTech Connect (OSTI)

The combined strong earthquake and super tsunami of 12 March 2011 at the Fukushima nuclear power plant imposed the most severe challenges ever experienced at such a facility. Information regarding the plant response and status remains uncertain, but it is clear that severe damage has been sustained, that the plant staff have responded creatively and that the offsite implications are unlikely to be seriously threatening to the health, if not the prosperity, of the surrounding population. Re-examination of the regulatory constraints of nuclear power will occur worldwide, and some changes are likely, particularly concerning reliance upon active systems for achieving critical safety functions and concerning treatments of used reactor fuel. Whether worldwide expansion of the nuclear power economy will be slowed in the long run is perhaps unlikely and worth discussion.

Golay, Michael (MIT) [MIT

2011-07-06T23:59:59.000Z

433

PNNL's Community Science & Technology Seminar Series Nuclear Power in a  

E-Print Network [OSTI]

PNNL's Community Science & Technology Seminar Series Nuclear Power in a Post-Fukushima World generated by nuclear power. What will the U.S. energy portfolio look like, and how will the energy demand is focused on longer- term operation of nuclear power plants, including measurements to detect

434

Ris9-R-609(EN) Simulation ofa PWR Power Plant  

E-Print Network [OSTI]

Ris9-R-609(EN) Simulation ofa PWR Power Plant for Process Control and Diagnosis Finn Ravnsbjerg Nielsen Risø National Laboratory, Roskilde, Denmark December 1991 #12;Simulation of a PWR Power Plant *^R a compute simulation of a simplified pressurized nuclear power plant model directed towards process control

435

Fragility Analysis Methodology for Degraded Structures and Passive Components in Nuclear Power Plants - Illustrated using a Condensate Storage Tank  

SciTech Connect (OSTI)

The Korea Atomic Energy Research Institute (KAERI) is conducting a five-year research project to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). The KAERI research project includes three specific areas that are essential to seismic probabilistic risk assessment (PRA): (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. Since 2007, Brookhaven National Laboratory (BNL) has entered into a collaboration agreement with KAERI to support its development of seismic capability evaluation technology for degraded structures and components. The collaborative research effort is intended to continue over a five year period. The goal of this collaboration endeavor is to assist KAERI to develop seismic fragility analysis methods that consider the potential effects of age-related degradation of structures, systems, and components (SSCs). The research results of this multi-year collaboration will be utilized as input to seismic PRAs. In the Year 1 scope of work, BNL collected and reviewed degradation occurrences in US NPPs and identified important aging characteristics needed for the seismic capability evaluations. This information is presented in the Annual Report for the Year 1 Task, identified as BNL Report-81741-2008 and also designated as KAERI/RR-2931/2008. The report presents results of the statistical and trending analysis of this data and compares the results to prior aging studies. In addition, the report provides a description of U.S. current regulatory requirements, regulatory guidance documents, generic communications, industry standards and guidance, and past research related to aging degradation of SSCs. In the Year 2 scope of work, BNL carried out a research effort to identify and assess degradation models for the long-term behavior of dominant materials that are determined to be risk significant to NPPs. Multiple models have been identified for concrete, carbon and low-alloy steel, and stainless steel. These models are documented in the Annual Report for the Year 2 Task, identified as BNL Report-82249-2009 and also designated as KAERI/TR-3757/2009. This report describes the research effort performed by BNL for the Year 3 scope of work. The objective is for BNL to develop the seismic fragility capacity for a condensate storage tank with various degradation scenarios. The conservative deterministic failure margin method has been utilized for the undegraded case and has been modified to accommodate the degraded cases. A total of five seismic fragility analysis cases have been described: (1) undegraded case, (2) degraded stainless tank shell, (3) degraded anchor bolts, (4) anchorage concrete cracking, and (5)a perfect combination of the three degradation scenarios. Insights from these fragility analyses are also presented.

Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y.; Kim, M.; Choi, I.

2010-06-30T23:59:59.000Z

436

The elements of nuclear power  

SciTech Connect (OSTI)

An introduction to the principles of nuclear fission power generation. Describes the physical processes which occur in a nuclear reactor and discusses the theory behind the calculations. Also covers heat transfer in reactors, thermodynamic power cycles, reactor operators, and radiation shielding. Material covered includes topics on the effects of nuclear radiation on humans, the safety of nuclear reactors and of those parts of the nuclear fuel cycle which deal with fuel element manufacture and the reprocessing of irradiated fuel.

Bennet, D.J.; Thomson, J.R.

1989-01-01T23:59:59.000Z

437

Resolving piping analysis issues to minimize impact on installation activities during refueling outage at nuclear power plants  

SciTech Connect (OSTI)

While it is required to maintain piping code compliance for all phases of installation activities during outages at a nuclear plant, it is equally essential to reduce challenges to the installation personnel on how plant modification work should be performed. Plant betterment activities that incorporate proposed design changes are continually implemented during the outages. Supporting analysis are performed to back these activities for operable systems. The goal is to reduce engineering and craft man-hours and minimize outage time. This paper outlines how plant modification process can be streamlined to facilitate construction teams to do their tasks that involve safety related piping. In this manner, installation can proceed by minimizing on the spot analytical effort and reduce downtime to support the proposed modifications. Examples are provided that permit performance of installation work in any sequence. Piping and hangers including the branch lines are prequalified and determined operable. The system is up front analyzed for all possible scenarios. The modification instructions in the work packages is flexible enough to permit any possible installation sequence. The benefit to this approach is large enough in the sense that valuable outage time is not extended and on site analytical work is not required.

Bhavnani, D. [Public Service Electric and Gas Co., Hancocks Bridge, NJ (United States)

1996-12-01T23:59:59.000Z

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