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Note: This page contains sample records for the topic "light water reactor" from the National Library of EnergyBeta (NLEBeta).
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1

Light Water Reactor Sustainability  

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

3 Light Water Reactor Sustainability Program ACCOMPLISHMENTS REPORT 2013 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

2

Light Water Reactor Sustainability  

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

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

3

Light Water Reactor Sustainability Newsletter  

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

hydraulics software RELAP-7 (which is under development in the Light Water Reactor Sustainability LWRS Program). A novel interaction between the probabilistic part (i.e., RAVEN)...

4

Light Water Reactor Sustainability Newsletter  

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

30-35, August 2012. Clayton, D. A. and M. S. Hileman, 2012, Light Water Reactor Sustainability Non-Destructive Evaluation for Concrete Research and Development Roadmap, ORNLTM-...

5

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

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

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1...

6

Light Water Reactor Sustainability (LWRS) Program  

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

Light Water Reactor Sustainability (LWRS) Program Login Instructions go here. User ID: Password: Log In Forgot your password?...

7

Light Water Reactor Sustainability Program Contact Information  

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

Program Organization LWRS Program Management Richard Reister Federal Project Director Light Water Reactor Deployment Office of Nuclear Energy U.S. Department of Energy...

8

Light-Water Breeder Reactor  

DOE Patents [OSTI]

Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

1972-06-20T23:59:59.000Z

9

Light Water Reactor Sustainability Newsletter  

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

and nuclear waste disposal. Dr. Corradini has extensive research experience in the phenomenology of beyond design basis Meet the New LWRS Program Pathway Lead accidents in light...

10

Zircaloy performance in light water reactors  

SciTech Connect (OSTI)

Zircaloy has been successfully used as the primary light water reactor (LWR) core structural material since its introduction in the early days of the US naval nuclear program. Its unique combination of low neutron absorption cross section, fabricability, mechanical strength, and corrosion resistance in water and steam near 300{degrees}C has resulted in remarkable reliability of operation of pressurized and boiling water reactor (PWR, BWR) fuel through the years. At present time, BWRs use Zircaloy-2 and PWRs use Zircaloy-4 for fuel cladding. In BWRs, both Zircaloy-2 and -4 have been successfully used for spacer grids and channels, and in PWRs Zircaloy-4 is used for spacer grids and control rod guide tubes. Performance of fuel rods has been excellent thus far. The current trend for utilities worldwide is to expect both higher fuel reliability in the future. Fuel suppliers have already achieved extended exposures in lead use assemblies, and have demonstrated excellent performance in all areas; therefore unsuspected problems are not likely to arise. However, as exposure and expectations continue to increase, Zircaloy is being taken toward the limits of its known capabilities. This paper reviews Zircaloy performance capabilities in areas related to environmentally affected microstructure, mechanical properties, corrosion resistance, and dimensional stability. The effects of radiation and reactor environment on each property is illustrated with data, micrographs, and analysis.

Adamson, R.B.; Cheng, B.C.; Kruger, R.M. [GE Nuclear Energy, Pleasanton, CA (United States)

1992-12-31T23:59:59.000Z

11

Light Water Reactor Sustainability Newsletter Kathryn McCarthy  

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

Technical Integration Office T he Light Water Reactor Sustainability (LWRS) Pro- gram Integrated Program Plan was released on January 31, 2012; it can be downloaded at https:...

12

Light Water Reactor Sustainability Newsletter Rebecca Smith-Kevern  

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

Rebecca Smith-Kevern Director, Office of Light Water Reactor Technologies. I am often asked why the Federal Government should fund a program that supports the continued operation...

13

Light Water Reactor Sustainability Program - Non-Destructive...  

Energy Savers [EERE]

for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants Light Water Reactor Sustainability Program - Non-Destructive Evaluation R&D Roadmap for...

14

Light Water Reactor Sustainability Newsletter By Rich Reister  

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

the safety, and extend the life of current reactors. The Light Water Re- actor Sustainability (LWRS) program is NE's principal means of achieving this objective. We have...

15

Light Water Reactor Sustainability Newsletter Thomas M. Rosseel  

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

Laboratory (ORNL), through the Department of Energy's (DOE) Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC (a...

16

Light Water Reactor Sustainability Newsletter By John Gaertner  

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

Year 2011 LWRS Program funding is very clear: "Regarding the Light Water Reactor Sustainability program, (Congress) expects a high cost share from industry." Cost sharing is...

17

Environmentally assisted cracking in light water reactors  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

1996-07-01T23:59:59.000Z

18

Light water reactor lower head failure analysis  

SciTech Connect (OSTI)

This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

1993-10-01T23:59:59.000Z

19

Safety of light water reactor fuel with silicon carbide cladding  

E-Print Network [OSTI]

Structural aspects of the performance of light water reactor (LWR) fuel rod with triplex silicon carbide (SiC) cladding - an emerging option to replace the zirconium alloy cladding - are assessed. Its behavior under accident ...

Lee, Youho

2013-01-01T23:59:59.000Z

20

Light Water Reactor Sustainability Newsletter Kathryn A. McCarthy  

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

S ome of the Light Water Reactor Sustainability (LWRS) Program managers have changed; therefore, I would like to provide a brief introduction to all of the LWRS program managers:...

Note: This page contains sample records for the topic "light water reactor" 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

Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels  

E-Print Network [OSTI]

Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

22

Rethinking the light water reactor fuel cycle  

E-Print Network [OSTI]

The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

Shwageraus, Evgeni, 1973-

2004-01-01T23:59:59.000Z

23

Issues affecting advanced passive light-water reactor safety analysis  

SciTech Connect (OSTI)

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01T23:59:59.000Z

24

Issues affecting advanced passive light-water reactor safety analysis  

SciTech Connect (OSTI)

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-08-01T23:59:59.000Z

25

Assessment of innovative fuel designs for high performance light water reactors  

E-Print Network [OSTI]

To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with ...

Carpenter, David Michael

2006-01-01T23:59:59.000Z

26

Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2  

SciTech Connect (OSTI)

The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

2002-09-01T23:59:59.000Z

27

The Consortium for Advanced Simulation of Light Water Reactors  

SciTech Connect (OSTI)

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

2011-10-01T23:59:59.000Z

28

Fuel Summary Report: Shippingport Light Water Breeder Reactor  

SciTech Connect (OSTI)

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01T23:59:59.000Z

29

Multi-Applications Small Light Water Reactor - NERI Final Report  

SciTech Connect (OSTI)

The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

2003-12-01T23:59:59.000Z

30

Cross section generation strategy for high conversion light water reactors  

E-Print Network [OSTI]

High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

Herman, Bryan R. (Bryan Robert)

2011-01-01T23:59:59.000Z

31

Accident Performance of Light Water Reactor Cladding Materials  

SciTech Connect (OSTI)

During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

Nelson, Andrew T. [Los Alamos National Laboratory

2012-07-24T23:59:59.000Z

32

Light Water Reactor Sustainability (LWRS) Program | Department of Energy  

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) "of EnergyEnergyENERGYWomen OwnedofDepartment ofJared Temanson -of Energy 1procedures,Light Water Reactor

33

Materials Inventory Database for the Light Water Reactor Sustainability Program  

SciTech Connect (OSTI)

Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

Kazi Ahmed; Shannon M. Bragg-Sitton

2013-08-01T23:59:59.000Z

34

Light Water Reactor Sustainability Constellation Pilot Project FY13 Summary Report  

SciTech Connect (OSTI)

Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY13.

R. Johansen

2013-09-01T23:59:59.000Z

35

Light Water Reactor Sustainability Constellation Pilot Project FY12 Summary Report  

SciTech Connect (OSTI)

Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY12.

R. Johansen

2012-09-01T23:59:59.000Z

36

Light-water-reactor safety research program. Quarterly progress report, July to September 1981  

SciTech Connect (OSTI)

Information is presented concerning environmentally assisted cracking in light water reactors; transient fuel response and fission-product release; and clad properties for code verification.

Not Available

1982-02-01T23:59:59.000Z

37

The impact of passive safety systems on desirability of advanced light water reactors  

E-Print Network [OSTI]

This work investigates whether the advanced light water reactor designs with passive safety systems are more desirable than advanced reactor designs with active safety systems from the point of view of uncertainty in the ...

Eul, Ryan C

2006-01-01T23:59:59.000Z

38

Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report  

SciTech Connect (OSTI)

A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

Lewis, M.R.

2000-01-11T23:59:59.000Z

39

Technologies for Upgrading Light Water Reactor Outlet Temperature  

SciTech Connect (OSTI)

Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

2013-07-01T23:59:59.000Z

40

Sustained Recycle in Light Water and Sodium-Cooled Reactors  

SciTech Connect (OSTI)

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

2010-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Light Water Reactor Sustainability Program Integrated Program Plan  

SciTech Connect (OSTI)

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

McCarthy, Kathryn A. [INL; Busby, Jeremy [ORNL; Hallbert, Bruce [INL; Bragg-Sitton, Shannon [INL; Smith, Curtis [INL; Barnard, Cathy [INL

2014-04-01T23:59:59.000Z

42

Light Water Reactor Sustainability Program Integrated Program Plan  

SciTech Connect (OSTI)

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

2012-01-01T23:59:59.000Z

43

Light Water Reactor Sustainability Program Integrated Program Plan  

SciTech Connect (OSTI)

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

2013-04-01T23:59:59.000Z

44

Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles  

SciTech Connect (OSTI)

The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

R. M. Ferrer; S. Bays; M. Pope

2008-04-01T23:59:59.000Z

45

Physics methods for calculating light water reactor increased performances  

SciTech Connect (OSTI)

The intensive use of light water reactors (LWRs) has induced modification of their characteristics and performances in order to improve fissile material utilization and to increase their availability and flexibility under operation. From the conceptual point of view, adequate methods must be used to calculate core characteristics, taking into account present design requirements, e.g., use of burnable poison, plutonium recycling, etc. From the operational point of view, nuclear plants that have been producing a large percentage of electricity in some countries must adapt their planning to the need of the electrical network and operate on a load-follow basis. Consequently, plant behavior must be predicted and accurately followed in order to improve the plant's capability within safety limits. The Belgonucleaire code system has been developed and extensively validated. It is an accurate, flexible, easily usable, fast-running tool for solving the problems related to LWR technology development. The methods and validation of the two computer codes LWR-WIMS and MICROLUX, which are the main components of the physics calculation system, are explained.

Vandenberg, C.; Charlier, A.

1988-11-01T23:59:59.000Z

46

Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report  

SciTech Connect (OSTI)

Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

R. Johansen

2011-09-01T23:59:59.000Z

47

The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel  

E-Print Network [OSTI]

Light water reactors (LWRs) are the world’s dominant nuclear reactor system. Uranium (U)-fuelled LWRs produce long-lived transuranic (TRU) isotopes. TRUs can be recycled in LWRs or fast reactors. The thermal neutron spectrum in LWRs is less suitable...

Lindley, Benjamin A.

2015-02-03T23:59:59.000Z

48

Multi-Application Small Light Water Reactor Final Report  

SciTech Connect (OSTI)

The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration, water desalination or district heating were not addressed directly in the economic analyses since these depend more on local conditions, demand and economy and can not be easily generalized. Current economic performance experience and available cost data were used. The preliminary cost estimate, based on a concept that could be deployed in less than a decade, is: (1) Net Electrical Output--1050 MWe; (2) Net Station Efficiency--23%; (3) Number of Power Units--30; (4) Nominal Plant Capacity Factor--95%; (5) Total capital cost--$1241/kWe; and (6) Total busbar cost--3.4 cents/kWh. The project includes a testing program that has been conducted at Oregon State University (OSU). The test facility is a 1/3-height and 1/254.7 volume scaled design that will operate at full system pressure and temperature, and will be capable of operation at 600 kW. The design and construction of the facility have been completed. Testing is scheduled to begin in October 2002. The MASLWR conceptual design is simple, safe, and economical. It operates at NSSS parameters much lower than for a typical PWR plant, and has a much simplified power generation system. The individual reactor modules can be operated as on/off units, thereby limiting operational transients to startup and shutdown. In addition, a plant can be built in increments that match demand increases. The ''pull and replace'' concept offers automation of refueling and maintenance activities. Performing refueling in a single location improves proliferation resistance and eliminates the threat of diversion. Design certification based on testing is simplified because of the relatively low cost of a full-scale prototype facility. The overall conclusion is that while the efficiency of the power generation unit is much lower (23% versus 30%), the reduction in capital cost due to simplification of design more than makes up for the increased cost of nuclear fuel. The design concept complies with the safety requirements and criteria. It also satisfies the goals for modularity, standard plant design, certification before construction, c

Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

2003-12-01T23:59:59.000Z

49

Fatigue and environmentally assisted cracking in light water reactors  

SciTech Connect (OSTI)

Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with {approx}300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289{degrees}C.

Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

1992-03-01T23:59:59.000Z

50

REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM  

SciTech Connect (OSTI)

The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

Nanstad, Randy K [ORNL; Odette, George Robert [UCSB

2010-01-01T23:59:59.000Z

51

Environmentally assisted cracking of light-water reactor materials  

SciTech Connect (OSTI)

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used.

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

1996-02-01T23:59:59.000Z

52

Materials science division light-water-reactor safety research program. Quarterly progress report, October - December 1981  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during October, November, and December 1981 on water-reactor-safety problems. The research and development areas covered are environmentally assisted cracking in light water reactors, transient fuel response and fission-product release, and clad properties for code verification.

Not Available

1982-05-01T23:59:59.000Z

53

Light-water-reactor safety research program. Quarterly progress report, April-June 1981  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during April, May, and June 1981 on water-reactor-safety problems. The research and development areas covered are transient fuel response and fission-product release and environmentally assisted cracking in light water reactors.

Not Available

1981-01-01T23:59:59.000Z

54

Materials Degradation in Light Water Reactors: Life After 60,???  

SciTech Connect (OSTI)

Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase susceptibility for most components and may introduce new degradation modes. While all components (except perhaps the reactor vessel) can be replaced, it may not be economically favorable. Therefore, understanding, controlling, and mitigating materials degradation processes are key priorities for reactor operation, power uprate considerations, and life extensions. This document is written to give an overview of some of the materials degradation issues that may be key for extend reactor service life. A detailed description of all the possible forms of degradation is beyond the scope of this short paper and has already been described in other documents (for example, the NUREG/CR-6923). The intent of this document is to present an overview of current materials issues in the existing reactor fleet and a brief analysis of the potential impact of extending life beyond 60 years. Discussion is presented in six distinct areas: (1) Reactor pressure vessel; (2) Reactor core and primary systems; (3) Reactor secondary systems; (4) Weldments; (5) Concrete; and (6) Modeling and simulations. Following each of these areas, some research thrust directions to help identify and mitigate lifetime extension issues are proposed. Note that while piping and cabling are important for extended service, these components are discussed in more depth in a separate paper. Further, the materials degradation issues associated with fuel cladding and fuel assemblies are not discussed in this section as these components are replaced periodically and will not influence the overall lifetime of the reactor.

Busby, Jeremy T [ORNL; Nanstad, Randy K [ORNL; Stoller, Roger E [ORNL; Feng, Zhili [ORNL; Naus, Dan J [ORNL

2008-04-01T23:59:59.000Z

55

INL/EXT-14-33257 Light Water Reactor Sustainability Program  

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

57 Light Water Reactor Sustainability Program 3D J-Integral Capability in Grizzly September 2014 DOE Office of Nuclear Energy DISCLAIMER This information was prepared as an account...

56

Assessment of light water reactor power plant cost and ultra-acceleration depreciation financing  

E-Print Network [OSTI]

Although in many regions of the U.S. the least expensive electricity is generated from light-water reactor (LWR) plants, the fixed (capital plus operation and maintenance) cost has increased to the level where the cost ...

El-Magboub, Sadek Abdulhafid.

57

EIS-0288: Production of Tritium in a Commercial Light Water Reactor  

Broader source: Energy.gov [DOE]

This Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS) evaluates the environmental impacts associated with producing tritium at one or more...

58

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

59

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems  

SciTech Connect (OSTI)

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

D. E. Shropshire

2009-01-01T23:59:59.000Z

60

Light-water-reactor safety materials engineering research programs. Quarterly progress report, January-March 1985. Volume 1  

SciTech Connect (OSTI)

This progress report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during January, February, and March 1985 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light-Water Reactors and Long-Term Embrittlement of Cast Duplex Stainless Steels in Light-Water-Reactor Systems. 42 refs.

Not Available

1986-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Light-water-reactor safety materials engineering research programs. Volume 3. Quarterly progress report, October-December 1984  

SciTech Connect (OSTI)

This progress report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during October, November, and December 1984 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light-Water Reactors and Long-Term Embrittlement of Cast Duplex Stainless Steels in Light-Water-Reactor Systems.

Not Available

1985-10-01T23:59:59.000Z

62

General features of direct-cycle, supercritical-pressure, light-water-cooled reactors  

SciTech Connect (OSTI)

The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR.

Oka, Y.; Koshizuka, S. [Univ. of Tokyo (Japan). Nuclear Engineering Research Lab.

1996-07-01T23:59:59.000Z

63

Materials Science Division light-water-reactor safety research program. Quarterly progress report, January-March 1982  

SciTech Connect (OSTI)

Information is presented concerning environmentally assisted cracking in light water reactors; transient fuel response and fission product release; and clad properties for code verification.

Shack, W.J.; Rest, J.; Kassner, T.F.

1982-10-01T23:59:59.000Z

64

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors  

SciTech Connect (OSTI)

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

2005-10-01T23:59:59.000Z

65

Light Water Reactor Safety Research Program. Semiannual report, April-September 1982  

SciTech Connect (OSTI)

This report documents progress made in Light Water Reactor Safety research conducted by Division 6441 in the period from April 1982 to September 1982. The programs conducted under investigation include Core Concrete Interactions, Core Melt-Coolant Interactions, Containment Emergency Sump Performance, the Hydrogen Program, and Combustible Gas in Containment Program. 50 references.

Berman, M.

1983-10-01T23:59:59.000Z

66

Light water reactor safety research program. Volume 12: quarterly report, Apr-Jun 79  

SciTech Connect (OSTI)

This report summarizes the progress of the Light Water Reactor Safety Research Program during the 2nd quarter of 1979. Specifically, the report summarizes progress in five major areas of research. They are: (1) the molten core/concrete interactions study; (2) steam explosion research phenomena; (3) statistical LOCA analysis; (4) UHI model development; (5) two-phase jet loads.

Berman, M.

1980-05-01T23:59:59.000Z

67

Evolutionary/advanced light water reactor data report  

SciTech Connect (OSTI)

The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

NONE

1996-02-09T23:59:59.000Z

68

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production  

SciTech Connect (OSTI)

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01T23:59:59.000Z

69

Light water reactor mixed-oxide fuel irradiation experiment  

SciTech Connect (OSTI)

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

Hodge, S.A.; Cowell, B.S. [Oak Ridge National Lab., TN (United States); Chang, G.S.; Ryskamp, J.M. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

1998-06-01T23:59:59.000Z

70

Consortium for Advanced Simulation of Light Water Reactors  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |Reactors The Consortium

71

Consortium for Advanced Simulation of Light Water Reactors  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |Reactors The

72

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |Reactors TheAdvanced

73

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |Reactors TheAdvancedHow

74

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |Reactors

75

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |ReactorsJournal and

76

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience |ReactorsJournal

77

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR Reactor Vessel

78

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR Reactor

79

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR ReactorNuclear

80

Overview of the US Department of Energy Light Water Reactor Sustainability Program  

SciTech Connect (OSTI)

The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

K. A. McCarthy; D. L. Williams; R. Reister

2012-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR MATERIALS"  

E-Print Network [OSTI]

309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR review; it is a book preview. Thirty years ago, "Fundamental Aspects of Nuclear Reactor Fuel Elements of nuclear fuels among other topics pertinent to the materials in the ensemble of the nuclear reactor

Motta, Arthur T.

82

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels  

E-Print Network [OSTI]

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

83

Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011  

SciTech Connect (OSTI)

The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

Not Listed

2011-11-01T23:59:59.000Z

84

Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community  

SciTech Connect (OSTI)

Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

2011-08-01T23:59:59.000Z

85

Fuel assembly for the production of tritium in light water reactors  

DOE Patents [OSTI]

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Cawley, William E. (Richland, WA); Trapp, Turner J. (Richland, WA)

1985-01-01T23:59:59.000Z

86

Fuel assembly for the production of tritium in light water reactors  

DOE Patents [OSTI]

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Cawley, W.E.; Trapp, T.J.

1983-06-10T23:59:59.000Z

87

Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [Argonne National Lab., IL (United States)] [and others

1997-10-01T23:59:59.000Z

88

Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [and others

1995-09-01T23:59:59.000Z

89

Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

1994-06-01T23:59:59.000Z

90

Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

1997-05-01T23:59:59.000Z

91

End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)  

SciTech Connect (OSTI)

Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

Richardson, K.D.

1987-10-01T23:59:59.000Z

92

Materials Science and Technology Division light-water-reactor safety research program: quarterly progress report, January-March 1983  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during January, February and March 1983 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, and Long-Term Embrittlement of Cast Duplex Stainless Steels in LWR Systems.

Not Available

1984-04-01T23:59:59.000Z

93

Materials Science Division light-water-reactor safety-research program. Quarterly progress report, April-June 1982. Volume 2  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during April, May, and June 1982 on water-reactor-safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, and Clad Properties for Code Verification.

Shack, W.J.; Rest, J.; Kassner, T.F.; Chung, H.M.; Claytor, T.N.; Kupperman, D.S.; Maiya, P.S.; Nichols, F.A.; Park, J.Y.; Ruther, W.E.; Yaggee, F.L.

1983-05-01T23:59:59.000Z

94

Materials Science Division light-water-reactor safety research program. Quarterly progress report, July-September 1982  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during July, August, and September 1982 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, Posttest Fuel Examination of the ORNL Fission Product Release Tests, and Examination of TMI-2 Fuel Specimens.

Shack, W.J.; Rest, J.; Kassner, T.F.; Neimark, L.A.; Chung, H.M.; Claytor, T.N.; Kupperman, D.S.; Maiya, P.S.; Nichols, F.A.; Park, J.Y.

1983-08-01T23:59:59.000Z

95

Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

1995-03-01T23:59:59.000Z

96

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels  

E-Print Network [OSTI]

1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

American Society for Testing and Materials. Philadelphia

2003-01-01T23:59:59.000Z

97

Materials Science and Technology Division, light-water-reactor safety research program. Quarterly progress report, October-December 1982  

SciTech Connect (OSTI)

The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, and Long-Term Embrittlement of Cast Duplex Stainless Steels in LWR Systems.

Shack, W.J.; Rest, J.; Kassner, T.F.; Ayrault, G.; Chopra, O.K.; Chung, H.M.; Kupperman, D.S.; Maiya, P.S.; Nichols, F.A.; Park, J.Y.

1983-11-01T23:59:59.000Z

98

Conceptual design of a pressure tube light water reactor with variable moderator control  

SciTech Connect (OSTI)

This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

Rachamin, R.; Fridman, E. [Reactor Safety Div., Inst. of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf, POB 51 01 19, 01314 Dresden (Germany); Galperin, A. [Dept. of Nuclear Engineering, Ben-Gurion Univ. of the Negev, POB 653, Beer Sheva 84105 (Israel)

2012-07-01T23:59:59.000Z

99

Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development  

SciTech Connect (OSTI)

Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

Shannon M. Bragg-Sitton

2013-09-01T23:59:59.000Z

100

Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)  

SciTech Connect (OSTI)

This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

Budd, W.A. (ed.)

1986-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

The Application of Structural Materials Data From the BN-350 Fast Reactor to Life Extension of Light Water Reactors  

SciTech Connect (OSTI)

This paper describes the results of investigations of 08Cr16Ni11Mo3 (AISI 316 steel analogue) austenitic stainless steel irradiated in BN-350 breeder reactor at irradiation conditions close to that for Light Water Reactor (LWR) Internals. The pores were found in 08Cr16Ni11Mo3 steel irradiated at temperature 280 deg. C up to rather low damage 1.3 dpa and with dose rate 3.9 x 10{sup -9} dpa/s. There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7-13 dpa at 302-311 deg. C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. There was observed an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the pores formation in the steel at low dose rates. (authors)

Romanenko, O.G. [Nuclear Technology Safety Center, Liza Chaikina 4, Almaty 050020 (Kazakhstan); Kislitsin, S.B.; Maksimkin, O.P. [Institute of Nuclear Physics, 1 Ibragimova St., Almaty, 050032 (Kazakhstan); Shiganakov, Sh.B.; Chumakov, Ye.V. [Kazakhstan Atomic Energy Committee, Liza Chaikina 4, Almaty (Kazakhstan); Dumchev, I.V. [MAEC Kazatomprom, Aktau, 130000 (Kazakhstan)

2006-07-01T23:59:59.000Z

102

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program  

SciTech Connect (OSTI)

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01T23:59:59.000Z

103

Stress corrosion cracking and crack tip characterization of Alloy X-750 in light water reactor environments  

E-Print Network [OSTI]

Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water ...

Gibbs, Jonathan Paul

2011-01-01T23:59:59.000Z

104

Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Light Water Reactor Environments  

E-Print Network [OSTI]

Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water ...

Gibbs, Jonathan Paul

105

Environmentally assisted cracking in light water reactors annual report January - December 2005.  

SciTech Connect (OSTI)

This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data, obtained in the pressurized water reactor environment, are presented on Ni-alloy welds prepared in the laboratory or obtained from the nozzle-to-pipe weld of the V. C. Summer reactor. The experimental CGRs under cyclic and constant load are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of these materials to environmentally enhanced cracking under a variety of loading conditions.

Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

2007-08-31T23:59:59.000Z

106

EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor (CLWR) Tritium Readiness Supplemental Environmental Impact Statement  

Broader source: Energy.gov [DOE]

This Supplemental EIS updates the environmental analyses in DOE’s 1999 EIS for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS). The CLWR EIS addressed the production of tritium in Tennessee Valley Authority reactors in Tennessee using tritium-producing burnable absorber rods.

107

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants  

SciTech Connect (OSTI)

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

Not Available

1993-05-13T23:59:59.000Z

108

Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels  

SciTech Connect (OSTI)

To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and the 18 samples from two boiling water reactor (BWR) fuel rods were examined. Their initial {sup 235}U enrichments and burnups were from 2.6 to 4.1% and from 4 to 50 GWd/t, respectively. One PWR fuel rod and one BWR fuel rod contained gadolinia as a burnable poison. The measurements for more than 40 nuclides of uranium, transuranium, and fission product elements were performed by destructive analysis using mass spectrometry, and alpha-ray and gamma-ray spectrometry. Burnup for each sample was determined by the {sup 148}Nd method. The analytical methods and the results as well as the related irradiation condition data are compiled as a complete benchmark data set.

Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kohno, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki [Japan Atomic Energy Research Institute (Japan)

2002-02-15T23:59:59.000Z

109

Microencapsulated Fuel Technology for Commercial Light Water and Advanced Reactor Application  

SciTech Connect (OSTI)

The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.

Terrani, Kurt A [ORNL; Snead, Lance Lewis [ORNL; Gehin, Jess C [ORNL

2012-01-01T23:59:59.000Z

110

Light Water Reactor Safety Research Program. Semiannual report, October 1983-March 1984  

SciTech Connect (OSTI)

This report describes the investigations and analyses conducted at Sandia National Laboratories, Albuquerque, in support of the Light Water Reactor Safety Research Program from October 1983 through March 1984. The Fuel-Coolant Interactions Study investigates the mechanism of concrete erosion by molten core materials, the nature and rate of generation of evolved gases, and the effects of fission-product release. The Hydrogen Behavior and Mitigative and Preventive Schemes Programs investigate the HECTR code for modeling hydrogen deflagration, and the Grand Gulf Igniter System II is being reviewed. All activities are continuing. 53 figs., 11 tabs.

Berman, M.

1986-02-01T23:59:59.000Z

111

Chemical aspects of pellet-cladding interaction in light water reactor fuel elements  

SciTech Connect (OSTI)

In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI.

Olander, D.R.

1982-01-01T23:59:59.000Z

112

Light water reactor safety research program. Quarterly report Jan-Mar 80  

SciTech Connect (OSTI)

The Molten Fuel Concrete Interactions (MFCI) study is comprised of experimental and analytical investigations of the chemical and physical phenomena associated with interactions between molten core materials and concrete. Such interactions are possible during hypothetical fuel-melt accidents in light water reactors (LWRs) when molten fuel and steel from the reactor core penetrate the pressure vessel and cascade onto the concrete substructure. The purpose of the MFCI study is to develop an understanding of these interactions suitable for risk assessment. Emphasis is placed on identifying and investigating the dominant interaction phenomena occurring between prototypic materials. The table of contents is the following: Molten fuel concrete interactions study; Steam explosion phenomena; Separate effects tests for TRAP code development; and Containment emergency sump performance.

Berman, M.

1980-09-01T23:59:59.000Z

113

Light-water-reactor safety research program. Quarterly progress report, July-September 1980  

SciTech Connect (OSTI)

A physically realistic description of fuel swelling and fission-gas release is needed to aid in predicting the behavior of fuel rods and fission gases under certain hypothetical light-water-reactor (LWR) accident conditions. To satisfy this need, a comprehensive computer-base model, the Steady-State and Transient Gas-Release and Swelling Subroutine (GRASS-SST), its faster-running version, FASTGRASS, and correlations based on analyses performed with GRASS-SST, PARAGRASS, are being developed at Argonne National Laboratory (ANL). This model is being incorporated into the Fuel-Rod Analysis Program (FRAP) code being developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL). The analytical effort is supported by a data base and correlations developed from characterization of irradiated LWR fuel and from out-of-reactor transient heating tests of irradiated commercial and experimental LWR fuel under a range of thermal conditions. 7 refs., 2 figs.

Massey, W.E.; Till, C.E.

1981-02-01T23:59:59.000Z

114

Core design study of a supercritical light water reactor with double row fuel rods  

SciTech Connect (OSTI)

An equilibrium core for supercritical light water reactor has been designed. A novel type of fuel assembly with dual rows of fuel rods between water rods is chosen and optimized to get more uniform assembly power distributions. Stainless steel is used for fuel rod cladding and structural material. Honeycomb structure filled with thermal isolation is introduced to reduce the usage of stainless steel and to keep moderator temperature below the pseudo critical temperature. Water flow scheme with ascending coolant flow in inner regions is carried out to achieve high outlet temperature. In order to enhance coolant outlet temperature, the radial power distributions needs to be as flat as possible through operation cycle. Fuel loading pattern and control rod pattern are optimized to flatten power distribution at inner regions. Axial fuel enrichment is divided into three parts to control axial power peak, which affects maximum cladding surface temperature. (authors)

Zhao, C.; Wu, H.; Cao, L.; Zheng, Y. [School of Nuclear Science and Technology, Xi'an Jiaotong Univ., No. 28, Xianning West Road, Xi'an, ShannXi, 710049 (China); Yang, J.; Zhang, Y. [China Nuclear Power Technology Research Inst., Yitian Road, ShenZhen, GuangDong, 518026 (China)

2012-07-01T23:59:59.000Z

115

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Charges Relating to Nuclear Reactor Safety," 1976, availablestudies of light-water nuclear reactor safety, emphasizingstudies of overall nuclear reactor safety have been

Nero, A.V.

2010-01-01T23:59:59.000Z

116

Modeling of the performance of weapons MOX fuel in light water reactors  

SciTech Connect (OSTI)

Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L. [Texas A and M Univ., College Station, TX (United States). Nuclear Engineering Dept.; Gellene, G.I. [Texas Tech Univ., Lubbock, TX (United States). Dept. of Chemistry and Biochemistry

1999-05-01T23:59:59.000Z

117

Adaptation of gas tagging for failed fuel identification in light water reactors  

SciTech Connect (OSTI)

This paper discusses experience with noble gas tagging and its adaptation to commercial reactors. It reviews the recent incidence of fuel failures in light water reactors, and methods used to identify failures, and concludes that the on-line technique of gas tagging could significantly augment present flux tilting, sipping and ultrasonic testing of assemblies. The paper describes calculations on tag gas stability in-reactor, and tag injection tests that were carried out collaboratively with Commonwealth Edison Company in the Byron-2 pressurized water reactor (P%a) and with Duke Power Company and Babcock and Wilcox Fuel Company in the Oconee-2 PWM. The tests gave information on: (a) noble gas concentration dynamics as the tag gases were dissolved in and eventually removed from subsystems of the RCS; and (b) the suitability of candidate Ar, Ne, Kr and Xe isotopes for tagging PWR fuel. It was found that the activity of Xe{sup 125} (the activation product of the tag isotope Xe{sup 124}) acted as a ``tag of a tag`` and tracked gas through the reactor; measured activities are being used to model gas movement in the RCS. Several interference molecules (trace contaminants normally present at sub-ppM concentrations in RCS samples) and entrained air in the RCS were found to affect mass spectrometer sensitivity for tag isotopes. In all instances the contaminants could be differentiated from the tag isotopes by operating the mass spectrometer at high resolution (2500). Similarly, it was possible to distinguish all the candidate tag gases against a high background of air. The test results suggested, however, that for routine analysis a high resolution static mass spectrometer will be preferable to the dynamic instrument used for the present analyses.

Lambert, J.D.B.; Gross, K.C.; Depiante, E.V. [Argonne National Lab., IL (United States); Callis, E.L. [Los Alamos National Lab., NM (United States); Egebrecht, P.M. [Commonwealth Edison Company, Downers Grove, IL (United States)

1996-03-01T23:59:59.000Z

118

Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors  

SciTech Connect (OSTI)

Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs.

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T. (Pacific Northwest Lab., Richland, WA (USA))

1989-11-01T23:59:59.000Z

119

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report  

SciTech Connect (OSTI)

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

Mac Donald, Philip Elsworth

2002-06-01T23:59:59.000Z

120

Feasibility Study of Supercritical Light Water Cooled Reactors for Electrical Power Production, 5th Quarterly Report, October - December 2002  

SciTech Connect (OSTI)

The overall objective of this project is to evaluate the feasibility of supercritical light water cooled reactors for electric power production. The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies for the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR that can also burn actinides. The project is organized into three tasks:

Philip MacDonald; Jacopo Buongiorno; Cliff Davis; J. Stephen Herring; Kevan Weaver; Ron Latanision; Bryce Mitton; Gary Was; Luca Oriani; Mario Carelli; Dmitry Paramonov; Lawrence Conway

2003-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" from the National Library of EnergyBeta (NLEBeta).
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121

Assessment of the use of extended burnup fuel in light water power reactors  

SciTech Connect (OSTI)

This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

1988-02-01T23:59:59.000Z

122

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel  

SciTech Connect (OSTI)

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Cowell, B.S.; Fisher, S.E.

1999-02-01T23:59:59.000Z

123

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J.; Smith, J. L.; Soppet, W. K.; Strain; R. V. (Energy Technology); ( APS-USR)

1999-10-01T23:59:59.000Z

124

Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design  

SciTech Connect (OSTI)

A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates that the proposed solutions to the investigated operating cycle length barriers are both feasible and consistent with sound design practice.

Professor Neill Todreas

2001-10-01T23:59:59.000Z

125

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors  

SciTech Connect (OSTI)

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding material both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

Rebak, Raul B. [General Electric] (ORCID:0000000280704475)

2014-12-30T23:59:59.000Z

126

Environmentally assisted cracking in light water reactors - annual report, January-December 2001.  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several heat treatment conditions.

Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

2003-06-01T23:59:59.000Z

127

Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)  

SciTech Connect (OSTI)

Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percent of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.

Campbell, W.R.; Giovengo, J.F.

1987-10-01T23:59:59.000Z

128

Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage  

SciTech Connect (OSTI)

The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

Bailey, W.J.

1987-11-01T23:59:59.000Z

129

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel  

SciTech Connect (OSTI)

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

2014-06-10T23:59:59.000Z

130

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel  

DOE Patents [OSTI]

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

2013-11-05T23:59:59.000Z

131

Light water reactor safety research program, quarterly report, July-September 1980. Volume 3  

SciTech Connect (OSTI)

The report covers research performed during July-September 1980 for the NRC Light Water Reactor Safety Research Program comprised of: (1) The Molten Fuel Concrete Interactions (MFCI) study of experimental and analytical investigations of the chemical and physical phenomena associated with interactions between molten core materials and concrete; (2) Steam Explosion Phenomena program to assess the probability and consequences of steam explosions during postulated meltdown accidents in LWRs; (3) Separate Effects Tests for TRAP Code Development investigating vapor pressures of fission-product species at elevated temperatures, chemical compound formation and reaction rates; (4) Containment Emergency Sump Performance (CESP) program to investigate the reliability of ECCS sumps; (5) Hydrogen Program designed to quantify the threat posed by hydrogen released during LWR accidents; and (6) Combustible Gas in Containment Program to study the generation of H2 from the corrosion of zinc and other materials located within LWR containment buildings.

Berman, M.

1981-04-01T23:59:59.000Z

132

Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project  

SciTech Connect (OSTI)

The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

1987-05-01T23:59:59.000Z

133

Analysis of assembly serial number usage in domestic light-water reactors  

SciTech Connect (OSTI)

Domestic light-water reactor (LWR) fuel assemblies are identified by a serial number that is placed on each assembly. These serial numbers are used as identifiers throughout the life of the fuel. The uniqueness of assembly serial numbers is important in determining their effectiveness as unambiguous identifiers. The purpose of this study is to determine what serial numbering schemes are used, the effectiveness of these schemes, and to quantify how many duplicate serial numbers occur on domestic LWR fuel assemblies. The serial numbering scheme adopted by the American National Standards Institute (ANSI) ensures uniqueness of assembly serial numbers. The latest numbering scheme adopted by General Electric (GE), was also found to be unique. Analysis of 70,971 fuel assembly serial numbers from permanently discharged fuel identified 11,948 serial number duplicates. Three duplicate serial numbers were found when analysis focused on duplication within the individual fuel inventory at each reactor site, but these were traced back to data entry errors and will be corrected by the Energy Information Administration (EIA). There were also three instances where the serial numbers used to identify assemblies used for hot cell studies differed from the serial numbers reported to the EIA. It is recommended that fuel fabricators and utilities adhere to the ANSI serial numbering scheme to ensure serial number uniqueness. In addition, organizations collecting serial number information, should request that all known serial numbers physically attached or associated with each assembly be reported and identified by the corresponding number scheme. 10 refs., 5 tabs.

Reich, W.J. (Oak Ridge National Lab., TN (USA)); Moore, R.S. (Automated Sciences Group, Inc., Oak Ridge, TN (USA))

1991-05-01T23:59:59.000Z

134

Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics  

SciTech Connect (OSTI)

The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

2014-02-01T23:59:59.000Z

135

Light-water-reactor safety research program. Quarterly progress report, January-March 1980  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.

Massey, W.E.; Kyger, J.A.

1980-08-01T23:59:59.000Z

136

Light-water-reactor safety research program: quarterly progress report, July-September, 1980  

SciTech Connect (OSTI)

The progress report summarizes the Argonne National Laboratory work performed during July, August, and September 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-product Release.

Massey, W.E.; Till, C.E.

1981-04-01T23:59:59.000Z

137

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production  

SciTech Connect (OSTI)

The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

2005-02-13T23:59:59.000Z

138

Radioactive Fission Product Release from Defective Light Water Reactor Fuel Elements  

SciTech Connect (OSTI)

Results are provided of the experimental investigation of radioactive fission product (RFP) release, i.e., krypton, xenon, and iodine radionuclides from fuel elements with initial defects during long-term (3 to 5 yr) irradiation under low linear power (5 to 12 kW/m) and during special experiments in the VK-50 vessel-type boiling water reactor.The calculation model for the RFP release from the fuel-to-cladding gap of the defective fuel element into coolant was developed. It takes into account the convective transport in the fuel-to-cladding gap and RFP sorption on the internal cladding surface and is in good agreement with the available experimental data. An approximate analytical solution of the transport equation is given. The calculation dependencies of the RFP release coefficients on the main parameters such as defect size, fuel-to-cladding gap, temperature of the internal cladding surface, and radioactive decay constant were analyzed.It is shown that the change of the RFP release from the fuel elements with the initial defects during long-term irradiation is, mainly, caused by fuel swelling followed by reduction of the fuel-to-cladding gap and the fuel temperature. The calculation model for the RFP release from defective fuel elements applicable to light water reactors (LWRs) was developed. It takes into account the change of the defective fuel element parameters during long-term irradiation. The calculation error according to the program does not exceed 30% over all the linear power change range of the LWR fuel elements (from 5 to 26 kW/m)

Konyashov, Vadim V.; Krasnov, Alexander M. [State Scientific Centre of Russian Federation-Research Institute of Atomic Reactors (Russian Federation)

2002-04-15T23:59:59.000Z

139

Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range on crack growth rates in air.

Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

2002-04-01T23:59:59.000Z

140

Light Water Reactor Sustainability Program A Reference Plan for Control Room Modernization: Planning and Analysis Phase  

SciTech Connect (OSTI)

The U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) program is collaborating with a U.S. nuclear utility to bring about a systematic fleet-wide control room modernization. To facilitate this upgrade, a new distributed control system (DCS) is being introduced into the control rooms of these plants. The DCS will upgrade the legacy plant process computer and emergency response facility information system. In addition, the DCS will replace an existing analog turbine control system with a display-based system. With technology upgrades comes the opportunity to improve the overall human-system interaction between the operators and the control room. To optimize operator performance, the LWRS Control Room Modernization research team followed a human-centered approach published by the U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 3, Human Factors Engineering Program Review Model (O’Hara et al., 2012), prescribes four phases for human factors engineering. This report provides examples of the first phase, Planning and Analysis. The three elements of Planning and Analysis in NUREG-0711 that are most crucial to initiating control room upgrades are: • Operating Experience Review: Identifies opportunities for improvement in the existing system and provides lessons learned from implemented systems. • Function Analysis and Allocation: Identifies which functions at the plant may be optimally handled by the DCS vs. the operators. • Task Analysis: Identifies how tasks might be optimized for the operators. Each of these elements is covered in a separate chapter. Examples are drawn from workshops with reactor operators that were conducted at the LWRS Human System Simulation Laboratory HSSL and at the respective plants. The findings in this report represent generalized accounts of more detailed proprietary reports produced for the utility for each plant. The goal of this LWRS report is to disseminate the technique and provide examples sufficient to serve as a template for other utilities’ projects for control room modernization.

Jacques Hugo; Ronald Boring; Lew Hanes; Kenneth Thomas

2013-09-01T23:59:59.000Z

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141

Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24  

SciTech Connect (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments.

Chopra, O.K.; Chung, H.M.; Gruber, E.E. [Argonne National Lab., IL (United States)] [and others

1998-04-01T23:59:59.000Z

142

Safeguards and security requirements for weapons plutonium disposition in light water reactors  

SciTech Connect (OSTI)

This paper explores the issues surrounding the safeguarding of the plutonium disposition process in support of the United States nuclear weapons dismantlement program. It focuses on the disposition of the plutonium by burning mixed oxide fuel in light water reactors (LWR) and addresses physical protection, material control and accountability, personnel security and international safeguards. The S and S system needs to meet the requirements of the DOE Orders, NRC Regulations and international safeguards agreements. Experience has shown that incorporating S and S measures into early facility designs and integrating them into operations provides S and S that is more effective, more economical, and less intrusive. The plutonium disposition safeguards requirements with which the US has the least experience are the implementation of international safeguards on plutonium metal; the large scale commercialization of the mixed oxide fuel fabrication; and the transportation to and loading in the LWRs of fresh mixed oxide fuel. It is in these areas where the effort needs to be concentrated if the US is to develop safeguards and security systems that are effective and efficient.

Thomas, L.L.; Strait, R.S. [Lawrence Livermore National Lab., CA (United States). Fission Energy and Systems Safety Program

1994-10-01T23:59:59.000Z

143

Insights for aging management of light water reactor components: Metal containments. Volume 5  

SciTech Connect (OSTI)

This report evaluates the available technical information and field experience related to management of aging damage to light water reactor metal containments. A generic aging management approach is suggested for the effective and comprehensive aging management of metal containments to ensure their safe operation. The major concern is corrosion of the embedded portion of the containment vessel and detection of this damage. The electromagnetic acoustic transducer and half-cell potential measurement are potential techniques to detect corrosion damage in the embedded portion of the containment vessel. Other corrosion-related concerns include inspection of corrosion damage on the inaccessible side of BWR Mark I and Mark II containment vessels and corrosion of the BWR Mark I torus and emergency core cooling system piping that penetrates the torus, and transgranular stress corrosion cracking of the penetration bellows. Fatigue-related concerns include reduction in the fatigue life (a) of a vessel caused by roughness of the corroded vessel surface and (b) of bellows because of any physical damage. Maintenance of surface coatings and sealant at the metal-concrete interface is the best protection against corrosion of the vessel.

Shah, V.N.; Sinha, U.P. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Smith, S.K. [Ogden Environmental and Energy Services, Southfield, MI (United States)

1994-03-01T23:59:59.000Z

144

Qualification Requirements of Guided Ultrasonic Waves for Inspection of Piping in Light Water Reactors  

SciTech Connect (OSTI)

Guided ultrasonic waves (GUW) are being increasingly used for both NDT and monitoring of piping. GUW offers advantages over many conventional NDE technologies due to the ability to inspect large volumes of piping components without significant removal of thermal insulation or protective layers. In addition, regions rendered inaccessible to more conventional NDE technologies may be more accessible using GUW techniques. For these reasons, utilities are increasingly considering the use of GUWs for performing the inspection of piping components in nuclear power plants. GUW is a rapidly evolving technology and its usage for inspection of nuclear power plant components requires refinement and qualification to ensure it is able to achieve consistent and acceptable levels of performance. This paper will discuss potential requirements for qualification of GUW techniques for the inspection of piping components in light water reactors (LWRs). The Nuclear Regulatory Commission has adopted ASME Boiler and Pressure Vessel Code requirements in Sections V, III, and XI for nondestructive examination methods, fabrication inspections, and pre-service and in-service inspections. A Section V working group has been formed to place the methodology of GUW into the ASME Boiler and Pressure Vessel Code but no requirements for technique, equipment, or personnel exist in the Code at this time.

Meyer, Ryan M.; Ramuhalli, Pradeep; Doctor, Steven R.; Bond, Leonard J.

2013-08-01T23:59:59.000Z

145

Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary  

SciTech Connect (OSTI)

Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

Shannon Bragg-Sitton

2014-02-01T23:59:59.000Z

146

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, October 1990--March 1991: Volume 13  

SciTech Connect (OSTI)

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties.

Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

1992-07-01T23:59:59.000Z

147

Light-Water-Reactor Safety Research Program. Quarterly progress report, October-December 1979  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during October, November, and December 1979 on water-reactor-safety problems. The research and development areas covered are: (1) Heat Transfer Coordination for LOCA Research Programs and (2) Transient Fuel Response and Fission-Product Release. 29 refs., 39 figs., 1 tab.

Massey, W.E.; Kyger, J.A.

1980-05-01T23:59:59.000Z

148

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM ADVANCED INSTRUMENTATION, INFORMATION, AND CONTROL SYSTEMS TECHNOLOGIES TECHNICAL PROGRAM PLAN FOR 2013  

SciTech Connect (OSTI)

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

Hallbert, Bruce; Thomas, Ken

2014-07-01T23:59:59.000Z

149

Light Water Reactor Sustainability Program Grizzly Year-End Progress Report  

SciTech Connect (OSTI)

The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. The reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i.e. late blooming phases) to become active later in life beyond the current operational experience. To develop a tool that can eventually serve a role in decision-making, it is clear that research and development must be perfomed at multiple scales. At the engineering scale, a multiphysics analysis code that can capture the thermomechanical response of the RPV under accident conditions, including detailed fracture mechanics evaluations of flaws with arbitrary geometry and orientation, is needed to assess whether the fracture toughness, as defined by the master curve, including the effects of embrittlement, is exceeded. At the atomistic scale, the fundamental mechanisms of degradation need to be understood, including the effects of that degradation on the relevant material properties. In addition, there is a need to better understand the mechanisms leading to the transition from ductile to brittle fracture through improved continuum mechanics modeling at the fracture coupon scale. Work is currently being conducted at all of these levels with the goal of creating a usable engineering tool informed by lower length-scale modeling. This report summarizes progress made in these efforts during FY 2013.

Benjamin Spencer; Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner; Marie Backman; Brian Wirth; Stephen Novascone; Jason Hales

2013-09-01T23:59:59.000Z

150

Materials Science and Technology Division Light-Water-Reactor Safety Research Program. Quarterly progress report, April-June 1983. Volume 2  

SciTech Connect (OSTI)

The progress report summarizes the Argonne National Laboratory work performed during April, May, and June 1983 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, and Long-Term Embrittlement of Cast Duplex Stainless Steels in LWR Systems.

Shack, W.J.

1984-06-01T23:59:59.000Z

151

Materials Science and Technology Division light-water-reactor safety research program. Quarterly progress report, July-September 1983. Volume 3  

SciTech Connect (OSTI)

This progress report summarizes the Argonne National Laboratory work performed during July, August, and September 1983 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors (reported elsewhere), Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, and Long-Term Embrittlement of Cast Duplex Stainless Steels in LWR Systems (reported elsewhere).

Not Available

1984-07-01T23:59:59.000Z

152

Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013  

SciTech Connect (OSTI)

Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

Idaho National Laboratory

2009-12-01T23:59:59.000Z

153

Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan  

SciTech Connect (OSTI)

This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

Yoshiaki Oka [Nuclear Engineering Research Laboratory, The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 112-0006 (Japan); Katsumi Yamada [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2004-07-01T23:59:59.000Z

154

Materials Science and Technology Division Light-Water-Reactor Safety Research Program. Volume 4. Quarterly progress report, October-December 1983  

SciTech Connect (OSTI)

The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, and Long-Term Embrittlement of Cast Duplex Stainless Steels in LWR Systems.

Not Available

1984-08-01T23:59:59.000Z

155

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels  

SciTech Connect (OSTI)

This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.

Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

1991-10-01T23:59:59.000Z

156

Reactor water cleanup system  

DOE Patents [OSTI]

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Gluntz, D.M.; Taft, W.E.

1994-12-20T23:59:59.000Z

157

Reactor water cleanup system  

DOE Patents [OSTI]

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

158

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL  

SciTech Connect (OSTI)

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-03-10T23:59:59.000Z

159

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors  

E-Print Network [OSTI]

Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced...

Billings, Jay Jay; Hull, S Forest; Lingerfelt, Eric J; Wojtowicz, Anna

2014-01-01T23:59:59.000Z

160

Extended-burnup LWR (light-water reactor) fuel: The amount, characteristics, and potential effects on interim storage  

SciTech Connect (OSTI)

The results of a study on extended-burnup, light-water reactor (LWR) spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory for the US Department of Energy (DOE). The purpose of the study was to collect and evaluate information on the status of in-reactor performance and integrity of extended-burnup LWR fuel and initiate the investigation of the effects of extending fuel burnup on the subsequent handling, interim storage, and other operations (e.g., rod consolidation and shipping) associated with the back end of the fuel cycle. The results of this study will aid DOE and the nuclear industry in assessing the effects on waste management of extending the useful in-reactor life of nuclear fuel. The experience base with extended-burnup fuel is now substantial and projections for future use of extended-burnup fuel in domestic LWRs are positive. The basic performance and integrity of the fuel in the reactor has not been compromised by extending the burnup, and the potential limitations for further extending the burnup are not severe. 104 refs., 15 tabs.

Bailey, W.J.

1989-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports  

SciTech Connect (OSTI)

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

NONE

1993-09-15T23:59:59.000Z

162

An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium  

SciTech Connect (OSTI)

An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

Wojtaszek, D.; Edwards, G. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ontario (Canada)

2013-07-01T23:59:59.000Z

163

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Volume 14, Semiannual report, April 1991--September 1991  

SciTech Connect (OSTI)

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWR`s); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from April 1991 through September 1991.

Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

1992-07-01T23:59:59.000Z

164

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Volume 15, Semiannual report: October 1991--March 1992  

SciTech Connect (OSTI)

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992.

Doctor, S.R.; Diaz, A.A.; Friley, J.R. [Pacific Northwest Lab., Richland, WA (United States)

1993-09-01T23:59:59.000Z

165

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. Semiannual report, April 1992--September 1992: Volume 16  

SciTech Connect (OSTI)

The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992.

Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

1993-11-01T23:59:59.000Z

166

Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors  

SciTech Connect (OSTI)

This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

2004-11-17T23:59:59.000Z

167

E-Print Network 3.0 - advanced light-water reactors Sample Search...  

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

of low-energy antineutrino detectors, together... Nuclear reactor safeguards and monitoring with ... Source: Gratta, Giorgio - Kavli Institute for Particle Astrophysics...

168

Advanced Light Water Reactor Plants System 80+{trademark} Design Certification Program. Annual progress report, October 1, 1992--September 30, 1993  

SciTech Connect (OSTI)

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW{sub t} (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment.

Not Available

1993-12-31T23:59:59.000Z

169

Light Water Reactor Sustainability Program Risk-Informed Safety Margins Characterization (RISMC) PathwayTechnical Program Plan  

SciTech Connect (OSTI)

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

Curtis Smith; Cristian Rabiti; Richard Martineau

2012-11-01T23:59:59.000Z

170

Acoustic Emission and Guided Ultrasonic Waves for Detection and Continuous Monitoring of Cracks in Light Water Reactor Components  

SciTech Connect (OSTI)

Acoustic emission (AE) and guided ultrasonic waves (GUW) are considered for continuous monitoring and detection of cracks in Light Water Reactor (LWR) components. In this effort, both techniques are applied to the detection and monitoring of fatigue crack growth in a full scale pipe component. AE results indicated crack initiation and rapid growth in the pipe, and significant GUW responses were observed in response to the growth of the fatigue crack. After initiation, the crack growth was detectable with AE for approximately 20,000 cycles. Signals associated with initiation and rapid growth where distinguished based on total rate of activity and differences observed in the centroid frequency of hits. An intermediate stage between initiation and rapid growth was associated with significant energy emissions, though few hits. GUW exhibit a nearly monotonic trend with crack length with an exception of measurements obtained at 41 mm and 46 mm.

Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep; Watson, Bruce E.; Cumblidge, Stephen E.; Doctor, Steven R.; Bond, Leonard J.

2012-06-28T23:59:59.000Z

171

Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods  

SciTech Connect (OSTI)

This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

1987-05-01T23:59:59.000Z

172

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts  

SciTech Connect (OSTI)

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water Reactor (PWR) assemblies. In addition to consideration of this 'naive' use of TRISO fuel in LWRs, several refined options are briefly examined and others are identified for further consideration including the use of advanced, high density fuel forms and larger kernel diameters and TRISO packing fractions. The combination of 800 {micro}m diameter kernels of 20% enriched UN and 50% TRISO packing fraction yielded reactivity sufficient to achieve comparable burnup to present-day PWR fuel.

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

2012-04-01T23:59:59.000Z

173

Importance of Delayed Neutrons on the Coupled Neutronic-Thermohydraulic Stability of a Natural Circulation Heavy Water-Moderated Boiling Light Water-Cooled Reactor  

SciTech Connect (OSTI)

The coupled neutronic-thermohydraulic stability characteristics of a natural circulation heavy water-moderated boiling light water-cooled reactor was investigated analytically considering the effects of prompt and delayed neutrons. For this purpose, the reactor considered is the Indian Advanced Heavy Water Reactor. The analytical model considers a point kinetics model for the neutron dynamics, a homogeneous two-phase flow model for the coolant thermal hydraulics, and a lumped heat transfer model for the fuel thermal dynamics. A higher mode of oscillation having a frequency much greater than the density-wave oscillation frequency was observed if prompt neutrons alone were considered. The occurrence of a higher mode of oscillation was found to be dependent on the concentration of delayed neutrons, the void reactivity coefficient, and the fuel time constant. The core inlet subcooling is found to have different effects on the decay ratio of the fundamental and higher modes of oscillations. The influences of void reactivity coefficient and fuel time constant on the fundamental and higher modes of oscillations were also found to be opposite in nature.

Nayak, A.K. [Bhaha Atomic Research Centre (India); Aritomi, M. [Tokyo Institute of Technology (Japan); Raj, V. Venkat [Bhaha Atomic Research Centre (India)

2001-07-15T23:59:59.000Z

174

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle  

SciTech Connect (OSTI)

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A., E-mail: sedov@dhtp.kial.ru; Subbotin, S. A.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15T23:59:59.000Z

175

A global approach of the representativity concept: Application on a high-conversion light water reactor MOX lattice case  

SciTech Connect (OSTI)

The development of new types of reactor and the increase in the safety specifications and requirements induce an enhancement in both nuclear data knowledge and a better understanding of the neutronic properties of the new systems. This enhancement is made possible using ad hoc critical mock-up experiments. The main difficulty is to design these experiments in order to obtain the most valuable information. Its quantification is usually made by using representativity and transposition concepts. These theories enable to extract some information about a quantity of interest (an integral parameter) on a configuration, but generally a posteriori. This paper presents a more global approach of this theory, with the idea of optimizing the representativity of a new experiment, and its transposition a priori, based on a multiparametric approach. Using a quadratic sum, we show the possibility to define a global representativity which permits to take into account several quantities of interest at the same time. The maximization of this factor gives information about all quantities of interest. An optimization method of this value in relation to technological parameters (over-clad diameter, atom concentration) is illustrated on a high-conversion light water reactor MOX lattice case. This example tackles the problematic of plutonium experiment for the plutonium aging and a solution through the optimization of both the over-clad and the plutonium content. (authors)

Santos, N. D.; Blaise, P.; Santamarina, A. [CEA, DEN/DER/SPRC Cadarache, F-13108 Saint Paul-lez-Durance (France)

2013-07-01T23:59:59.000Z

176

Review of experiments to evaluate the ability of electrical heater rods to simulate nuclear fuel rod behavior during postulated loss-of-coolant accidents in light water reactors  

SciTech Connect (OSTI)

Issues related to using electrical fuel rod simulators to simulate nuclear fuel rod behavior during postulated loss-of-coolant accident (LOCA) conditions in light water reactors are summarized. Experimental programs which will provide a data base for comparing electrical heater rod and nuclear fuel rod LOCA responses are reviewed.

McPherson, G D; Tolman, E L

1980-01-01T23:59:59.000Z

177

Evaluation of weapons-grade mixed oxide fuel performance in U.S. Light Water Reactors using COMETHE 4D release 23 computer code  

E-Print Network [OSTI]

The COMETHE 4D Release 23 computer code was used to evaluate the thermal, chemical and mechanical performance of weapons-grade MOX fuel irradiated under U.S. light water reactor typical conditions. Comparisons were made to and UO? fuels exhibited...

Bellanger, Philippe

2012-06-07T23:59:59.000Z

178

Shielding analysis for the 300 area light water reactor spent nuclear fuel within a modified multi-canister overpack canister in a modified multi-canister overpack cask  

SciTech Connect (OSTI)

Spent light water reactor fuel is to be moved out of the 324 Building. It is anticipated that intact fuel assemblies will be loaded in a modified Multi-Canister Overpack Canister, which in turn will be placed in an Overpack Transportation Cask. An estimate of gamma ray dose rates from a transportation cask is desired.

Gedeon, S.R.

1997-04-11T23:59:59.000Z

179

Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL  

SciTech Connect (OSTI)

The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.

Michael A. Pope; Samuel E. Bays; S. Piet; R. Ferrer; Mehdi Asgari; Benoit Forget

2009-05-01T23:59:59.000Z

180

Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study  

SciTech Connect (OSTI)

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

2012-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report  

SciTech Connect (OSTI)

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

Philip E. MacDonald

2003-09-01T23:59:59.000Z

182

Silicon carbide performance as cladding for advanced uranium and thorium fuels for light water reactors  

E-Print Network [OSTI]

There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 ...

Sukjai, Yanin

2014-01-01T23:59:59.000Z

183

Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials  

SciTech Connect (OSTI)

The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.

Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

1995-07-01T23:59:59.000Z

184

Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN  

SciTech Connect (OSTI)

The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

2013-09-01T23:59:59.000Z

185

Human-In-The-Loop Simulation in Support of Long-Term Sustainability of Light Water Reactors  

SciTech Connect (OSTI)

Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration. The NPP owners and operators realize that this analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are needed to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II&C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation’s NPPs.

Bruce P Hallbert

2015-01-01T23:59:59.000Z

186

Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining  

SciTech Connect (OSTI)

A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl – 1 wt% Li2O at 650 °C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 °C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

S. D. Herrmann; S. X. Li

2010-09-01T23:59:59.000Z

187

Light Water Reactor Fuel Cladding Research and Testing | ornl.gov  

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 MayAtmospheric Optical Depth7-1D: VegetationEquipment Surfaces and Interfaces Sample6, 2011 LawrenceEfeedstocks and the climateLife a9Light

188

SEIS for the Production of Tritium in a Commercial Light Water Reactor |  

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 Administration the1 -the Mid-Infrared0 ResourceAwards SAGE Awards ,# , onLightThePrices andSEE HOW

189

Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method  

SciTech Connect (OSTI)

For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

2007-11-01T23:59:59.000Z

190

Fracture mechanics models developed for piping reliability assessment in light water reactors: piping reliability project  

SciTech Connect (OSTI)

The efforts concentrated on modifications of the stratified Monte Carlo code called PRAISE (Piping Reliability Analysis Including Seismic Events) to make it more widely applicable to probabilistic fracture mechanics analysis of nuclear reactor piping. Pipe failures are considered to occur as the result of crack-like defects introduced during fabrication, that escape detection during inspections. The code modifications allow the following factors in addition to those considered in earlier work to be treated: other materials, failure criteria and subcritical crack growth characteristic; welding residual and vibratory stresses; and longitudinal welds (the original version considered only circumferential welds). The fracture mechanics background for the code modifications is included, and details of the modifications themselves provided. Additionally, an updated version of the PRAISE user's manual is included. The revised code, known as PRAISE-B was then applied to a variety of piping problems, including various size lines subject to stress corrosion cracking and vibratory stresses. Analyses including residual stresses and longitudinal welds were also performed.

Harris, D.O.; Lim, E.Y.; Dedhia, D.D.; Woo, H.H.; Chou, C.K.

1982-06-01T23:59:59.000Z

191

Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors  

SciTech Connect (OSTI)

Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

1994-04-01T23:59:59.000Z

192

Light-water-reactor safety fuel systems research programs. Quarterly progress report, January-March 1985. Volume 1  

SciTech Connect (OSTI)

This progress report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during January, February, and March 1985 on water reactor safety problems related to fuel and cladding. The research and development areas covered are Transient Fuel Response and Fission Product Release and Clad Properties for Code Verification. 15 refs.

Not Available

1986-01-01T23:59:59.000Z

193

Light-water-reactor safety fuel systems research programs. Quarterly progress report, January-March 1984. [Fuel and cladding problems  

SciTech Connect (OSTI)

This progress report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during January, February, and March 1984 on water reactor safety problems related to fuel and cladding. The research and development areas covered are Transient Fuel Response and Fission Product Release and Clad Properties for Code Verification.

Not Available

1984-09-01T23:59:59.000Z

194

Light-water-reactor safety fuel systems research programs. Quarterly progress report, July-September 1984. Volume 3  

SciTech Connect (OSTI)

This progress report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during July, August, and September 1984 on water reactor safety problems related to fuel and cladding. The research and development areas covered are Transient Fuel Response and Fission Product Release and Clad Properties for Code Verification. 17 refs., 23 figs., 5 tabs.

Not Available

1985-04-01T23:59:59.000Z

195

Light-water-reactor safety fuel systems research programs. Quarterly progress report, April-June 1984. Volume 2  

SciTech Connect (OSTI)

This progress report report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during April, May, and June 1984 on water reactor safety problems related to fuel and cladding. The research and development areas covered are Transient Fuel Response and Fission Product Release and Clad Properties for Code Verification.

Not Available

1985-02-01T23:59:59.000Z

196

Light-water-reactor safety fuel systems research programs. Quarterly progress report, October-December 1984. Volume 4  

SciTech Connect (OSTI)

This progress report summarizes work performed by the Materials Science and Technology Division of Argonne National Laboratory during October, November, and December 1984 on water reactor safety problems related to fuel and cladding. The research and development areas covered are Transient Fuel Response and Fission Product Release and Clad Properties for Code Verification. 30 refs., 23 figs., 2 tabs.

Not Available

1985-08-01T23:59:59.000Z

197

Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors  

SciTech Connect (OSTI)

The overall goal of the project is to determine the mechanism of irradiation assisted stress corrosion cracking (IASCC). IASCC has been linked to hardening, microstructural and microchemical changes during irradiation. Unfortunately, all of these changes occur simultaneously and at similar rates during irradiation, making attribution of IASCC to any one of these features nearly impossible to determine. The strategy set forth in this project is to develop means to separate microstructural from microchemical changes to evaluate each separately for their effect on IASCC. In the first part, post irradiation annealing (PIA) treatments are used to anneal the irradiated microstructure, leaving only radiation induced segregation (RIS) for evaluation for its contribution to IASCC. The second part of the strategy is to use low temperature irradiation to produce a radiation damage dislocation loop microstructure without radiation induced segregation in order to evaluate the effect of the dislocation microstructure alone. A radiation annealing model was developed based on the elimination of dislocation loops by vacancy absorption. Results showed that there were indeed, time-temperature annealing combinations that leave the radiation induced segregation profile largely unaltered while the dislocation microstructure is significantly reduced. Proton irradiation of 304 stainless steel irradiated with 3.2 MeV protons to 1.0 or 2.5 dpa resulted in grain boundary depletion of chromium and enrichment of nickel and a radiation damaged microstructure. Post irradiation annealing at temperatures of 500 ? 600°C for times of up to 45 min. removed the dislocation microstructure to a greater degree with increasing temperatures, or times at temperature, while leaving the radiation induced segregation profile relatively unaltered. Constant extension rate tensile (CERT) experiments in 288°C water containing 2 ppm O2 and with a conductivity of 0.2 mS/cm and at a strain rate of 3 x 10-7 s-1 showed that the IASCC susceptibility, as measured by the crack length per unit strain, decreased with very short anneals and was almost completely removed by an anneal at 500°C for 45 min. This annealing treatment removed about 15% of the dislocation microstructure and the irradiation hardening, but did not affect the grain boundary chromium depletion or nickel segregation, nor did it affect the grain boundary content of other minor impurities. These results indicate that RIS is not the sole controlling feature of IASCC in irradiated stainless steels in normal water chemistry. The isolation of the irradiated microstructure was approached using low temperature irradiation or combinations of low and high temperature irradiations to achieve a stable, irradiated microstructure without RIS. Experiments were successful in achieving a high degree of irradiation hardening without any evidence of RIS of either major or minor elements. The low temperature irradiations to doses up to 0.3 dpa at T<75°C were also very successful in producing hardening to levels considerably above that for irradiations conducted under nominal conditions of 1 dpa at 360°C. However, the microstructure consisted of an extremely fine dispersion of defect clusters of sizes that are not resolvable by either transmission electron microscopy (TEM) or small angle x-ray scattering (SAXS). The microstructure was not stable at the 288°C IASCC test temperature and resulted in rapid reduction of hardening and presumably, annealing of the defect clusters at this temperature as well. Nevertheless, the annealing studies showed that treatments that resulted in significant decreases in the hardening produced small changes in the dislocation microstructure that were confined to the elimination of the finest of loops (~1 nm). These results substantiate the importance of the very fine defect microstructure in the IASCC process. The results of this program provide the first definitive evidence that RIS is not the sole controlling factor in the irradiation assisted stress corrosion cracking of austenitic stain

Gary S. Was; Michael Atzmon; Lumin Wang

2003-04-28T23:59:59.000Z

198

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report  

SciTech Connect (OSTI)

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

Mac Donald, Philip Elsworth

2002-09-01T23:59:59.000Z

199

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors  

SciTech Connect (OSTI)

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01T23:59:59.000Z

200

Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports  

SciTech Connect (OSTI)

This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

Not Available

1994-01-15T23:59:59.000Z

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201

E-Print Network 3.0 - argonne heavy water reactor Sample Search...  

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

electrolysis LHV Low heating value LWR Light water reactor MHR Modular helium reactor Q Heat SOEC Solid oxide... electrolysis cell SOFC Solid oxide fuel cell SCWR Super critical...

202

NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669  

SciTech Connect (OSTI)

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1994-08-01T23:59:59.000Z

203

NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669  

SciTech Connect (OSTI)

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1994-08-01T23:59:59.000Z

204

Hydrogen and water reactor safety: proceedings  

SciTech Connect (OSTI)

Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

Not Available

1982-01-01T23:59:59.000Z

205

SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4  

SciTech Connect (OSTI)

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

Hagrman, D.T. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

1995-06-01T23:59:59.000Z

206

Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)  

SciTech Connect (OSTI)

The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

Mac Donald, Philip Elsworth

2002-09-01T23:59:59.000Z

207

A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors  

SciTech Connect (OSTI)

This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of a small set of comprehensive event trees and fault trees and recommendation for future work.

S. Khericha

2011-06-01T23:59:59.000Z

208

Possible effects of UO/sub 2/ oxidation on light water reactor spent fuel performance in long-term geologic disposal  

SciTech Connect (OSTI)

Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO/sub 2/) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO/sub 2/ oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented.

Almassy, M.Y.; Woodley, R.E.

1982-08-01T23:59:59.000Z

209

TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis  

E-Print Network [OSTI]

The accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics/thermal-hydraulics codes have been developed for this ...

Griggs, D. P.

1984-01-01T23:59:59.000Z

210

Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data  

SciTech Connect (OSTI)

Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

J. W. Sterbentz

1999-08-01T23:59:59.000Z

211

Reduction of the Radiotoxicity of Spent Nuclear Fuel Using a Two-Tiered System Comprising Light Water Reactors and Accelerator-Driven Systems  

SciTech Connect (OSTI)

Two main issues regarding the disposal of spent nuclear fuel from nuclear reactors in the United States in the geological repository Yucca Mountain are: (1) Yucca Mountain is not designed to hold the amount of fuel that has been and is proposed to be generated in the next few decades, and (2) the radiotoxicity (i.e., biological hazard) of the waste (particularly the actinides) does not decrease below that of natural uranium ore for hundreds of thousands of years. One solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation; there are advantages and disadvantages associated with each. By using existing Light Water Reactors (LWRs) to burn a majority of the plutonium in spent nuclear fuel and Accelerator-Driven Systems (ADSs) to transmute the remainder of the actinides, the benefits of each type of system can be realized. The transmutation process then becomes more efficient and less expensive. This research searched for the best combination of LWRs with multiple recycling of plutonium and ADSs to transmute spent nuclear fuel from past and projected nuclear activities (assuming little growth of nuclear energy). The neutronic design of each system is examined in detail although thermal hydraulic performance would have to be considered before a final system is designed. The results are obtained using the Monte Carlo burnup code Monteburns, which has been successfully benchmarked for MOX fuel irradiation and compared to other codes for ADS calculations. The best combination of systems found in this research includes 41 LWRs burning mixed oxide fuel with two recycles of plutonium ({approx}40 years operation each) and 53 ADSs to transmute the remainder of the actinides from spent nuclear fuel over the course of 60 years of operation.

H.R. Trellue

2003-06-01T23:59:59.000Z

212

Optimization of hydride fueled pressurized water reactor cores  

E-Print Network [OSTI]

This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

Shuffler, Carter Alexander

2004-01-01T23:59:59.000Z

213

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies  

SciTech Connect (OSTI)

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation. As an initial step in accomplishing this effort, the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies was held March 20–21, 2009, in Columbus, Ohio, to enable industry stakeholders and researchers in identification of the nuclear industry’s needs in the areas of future I&C technologies and corresponding technology gaps and research capabilities. Approaches for collaboration to bridge or fill the technology gaps were presented and R&D activities and priorities recommended. This report documents the presentations and discussions of the workshop and is intended to serve as a basis for the plan under development to achieve the goals of the I&C research pathway.

Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

2009-08-01T23:59:59.000Z

214

Investigation of the use of nanofluids to enhance the In-Vessel Retention capabilities of Advanced Light Water Reactors  

E-Print Network [OSTI]

Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, ...

Hannink, Ryan Christopher

2007-01-01T23:59:59.000Z

215

A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle  

SciTech Connect (OSTI)

At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

Fishbone, L.G.; Higinbotham, W.A.

1986-06-01T23:59:59.000Z

216

Light Water Reactor Sustainability Newsletter  

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

the operation of commercial nuclear power plants require conservative mar- gins of fracture toughness for the RPV materials, both during normal operation and under accident...

217

Light Water Reactor Sustainability Newsletter  

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

The small volume required for such analysis is beneficial for correlating with the small-scale mechanical testing currently being investigated. Further studies on the...

218

Stability analysis of supercritical water cooled reactors  

E-Print Network [OSTI]

The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

2005-01-01T23:59:59.000Z

219

Feasibility study on the thorium fueled boiling water breeder reactor  

SciTech Connect (OSTI)

The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

PetrusTakaki, N. [Dept. of Applied Science, Tokai Univ., Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan)

2012-07-01T23:59:59.000Z

220

E-Print Network 3.0 - advanced light reactor Sample Search Results  

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

2 3 4 5 > >> Page: << < 1 2 3 4 5 > >> 21 Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Summary: at the end of 2000.2 Most are light water...

Note: This page contains sample records for the topic "light water reactor" 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

Heavy Water Test Reactor Dome Removal  

SciTech Connect (OSTI)

A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

None

2011-01-01T23:59:59.000Z

222

Evaluation of the tritium content in light water reactor control and absorber rods to obtain data for the fuel cycle backend  

SciTech Connect (OSTI)

Tritium inventories and tritium distribution have been determined in boron glass absorber rods discharged from a pressurized water reactor first-cycle core and in spent boron carbide (B/sub 4/C) control rods from a boiling water reactor. The total tritium inventory in the boron glass absorber rods from the Stade nuclear reactor amounts to approx. =8.0 x 10/sup 10/ Bq (2.2 Ci) per rod. Of this, 99.6% was fixed in the boron glass itself and 0.4% in the Al/sub 2/O/sub 3/ pellets. The 4 x 10/sup -3/% fractions in the tube cladding and support pipe and the 1 x 10/sup -2/% fraction in the fill gas accounted for an insignificant part of the total tritium inventory of the rod. This experimentally determined tritium inventory was a factor of 5 larger than that suggested by the calculated estimate. The discrepancy between analyzed and calculated values can be explained by tritium formation from lithium impurities in the boron glass, where a 30-ppm lithium content would be adequate for this tritium inventory to be generated by the reaction /sup 6/Li(n,..cap alpha..)/sup 3/H. Evaluation of the B/sub 4/C control rods from the Lingen nuclear reactor after 3 yr of operation gave a 3.2 x 10/sup 10/Bq(0.85-Ci)tritium inventory per B/sub 4/C rod, while the total tritium inventory for a control rod assembly containing 60 B/sub 4/C rods was approx. =1.9 x 10/sup 12/ Bq (50 Ci). The tritium generated was essentially bound 100% in the B/sub 4/C, since the hulls contained only 6 x 10/sup -3/% and the fill gas only 2 x 10/sup -4/%.

Bleier, A.; Neeb, K.H.; Gelfort, E.; Mischke, J.

1986-08-01T23:59:59.000Z

223

Review of High Temperature Water and Steam Cooled Reactor Concepts  

SciTech Connect (OSTI)

This review summarizes design concepts of supercritical-pressure water cooled reactors (SCR), nuclear superheaters and steam cooled fast reactors from 1950's to the present time. It includes water moderated supercritical steam cooled reactor, SCOTT-R and SC-PWR of Westinghouse, heavy water moderated light water cooled SCR of GE, SCLWR and SCFR of the University of Tokyo, B-500SKDI of Kurchatov Institute, CANDU -X of AECL, nuclear superheaters of GE, subcritical-pressure steam cooled FBR of KFK and B and W, Supercritical-pressure steam cooled FBR of B and W, subcritical-pressure steam cooled high converter by Edlund and Schultz and subcritical-pressure water-steam cooled FBR by Alekseev. This paper is prepared based on the previous review of SCR2000 symposium, and some author's comments are added. (author)

Oka, Yoshiaki [Nuclear Engineering Research Laboratory, The University of Tokyo, 3-1, Hongo 7-Chome, Bunkyo-ku (Japan)

2002-07-01T23:59:59.000Z

224

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report  

SciTech Connect (OSTI)

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.

Philip E. MacDonald

2005-01-01T23:59:59.000Z

225

TA-2 Water Boiler Reactor Decommissioning Project  

SciTech Connect (OSTI)

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

Durbin, M.E. (ed.); Montoya, G.M.

1991-06-01T23:59:59.000Z

226

Boiling water reactor control rod  

SciTech Connect (OSTI)

This patent describes a control rod for a boiling water type nuclear power reactor, the improvement comprising: (a) an elongated central stem defining a longitudinally extending internal central gas plenum; (b) blades connected to and extending along and radially outward from the stem, each blade including an elongated body portion extending along the stem and terminating in an end tip portion; (c) means defining a series of internal cavities in each of the blades, the cavities being arranged in columns and rows across the length and width of the body and tip portions of the blade; (d) pellets of neutron absorbing material, each disposed within one of the cavities with each of the cavities being oversized in relation to the size of the pellet disposed therein to allow extra space for swelling of the pellet. The cavities and the pellets disposed therein are arranged to define a longer, constant worth section generally coextensive with the body portion of the blade and a shorter, reduced worth section generally coextensive with the end tip portion of the blade; and (e) means defined within each blade communicating each of the cavities with the central gas plenum for allowing any gases generated by irradiation of the pellets to expand from the cavities into the plenum.

Wilson, J.F.; Doshi, P.K.

1986-12-23T23:59:59.000Z

227

Columbia Water and Light- HVAC and Lighting Efficiency Rebates  

Broader source: Energy.gov [DOE]

Columbia Water and Light (CWL) offers rebates to its commercial and industrial customers for the purchase of high efficiency HVAC installations and efficient lighting. Incentives for certain...

228

Self-Sustaining Thorium Boiling Water Reactors  

E-Print Network [OSTI]

A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar ...

Ganda, Francesco

229

Automatic reactor power control for a pressurized water reactor  

SciTech Connect (OSTI)

An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

1993-05-01T23:59:59.000Z

230

DOE plutonium disposition study: Analysis of existing ABB-CE Light Water Reactors for the disposition of weapons-grade plutonium. Final report  

SciTech Connect (OSTI)

Core reactivity and basic fuel management calculations were conducted on the selected reactors (with emphasis on the System 80 units as being the most desirable choice). Methods used were identical to those reported in the Evolutionary Reactor Report. From these calculations, the basic mission capability was assessed. The selected reactors were studied for modification, such as the addition of control rod nozzles to increase rod worth, and internals and control system modifications that might also be needed. Other system modifications studied included the use of enriched boric acid as soluble poison, and examination of the fuel pool capacities. The basic geometry and mechanical characteristics, materials and fabrication techniques of the fuel assemblies for the selected existing reactors are the same as for System 80+. There will be some differences in plutonium loading, according to the ability of the reactors to load MOX fuel. These differences are not expected to affect licensability or EPA requirements. Therefore, the fuel technology and fuel qualification sections provided in the Evolutionary Reactor Report apply to the existing reactors. An additional factor, in that the existing reactor availability presupposes the use of that reactor for the irradiation of Lead Test Assemblies, is discussed. The reactor operating and facility licenses for the operating plants were reviewed. Licensing strategies for each selected reactor were identified. The spent fuel pool for the selected reactors (Palo Verde) was reviewed for capacity and upgrade requirements. Reactor waste streams were identified and assessed in comparison to uranium fuel operations. Cost assessments and schedules for converting to plutonium disposition were estimated for some of the major modification items. Economic factors (incremental costs associated with using weapons plutonium) were listed and where possible under the scope of work, estimates were made.

Not Available

1994-06-01T23:59:59.000Z

231

Operational control of boiling water reactor stability  

SciTech Connect (OSTI)

Boiling water reactor cores are susceptible to instabilities, which generate power oscillations. Specific reactor operating practices can provide a mechanism for control of the instability phenomenon. An axial separation of the core into a single-phase region and a two-phase region resolves the influence of axial flux shapes on core stability. This separation provides the means to derive a core stability control that ensures significant reactor stability margin. The control is achieved by maintaining the core average bulk coolant saturation elevation above a predetermined axial plane. The control can be reliably and efficiently implemented during reactor operations. Analysis demonstrates that variations in parameters important to stability have only secondary influences on stability margin when the control is in effect. Actual plant experience with a large commercial boiling water reactor confirms the capabilities of this stability control in an operational setting.

Mowry, C.M. [PECO Energy, Wayne, PA (United States); Nir, I. [Entergy Operations, Jackson, MS (United States); Newkirk, D.W. [GE Nuclear Energy, San Jose, CA (United States)

1995-03-01T23:59:59.000Z

232

Steam turbine: Alternative emergency drive for the secure removal of residual heat from the core of light water reactors in ultimate emergency situation  

SciTech Connect (OSTI)

In 2011 the nuclear power generation has suffered an extreme probation. That could be the meaning of what happened in Fukushima Nuclear Power Plants. In those plants, an earthquake of 8.9 on the Richter scale was recorded. The quake intensity was above the trip point of shutting down the plants. Since heat still continued to be generated, the procedure to cooling the reactor was started. One hour after the earthquake, a tsunami rocked the Fukushima shore, degrading all cooling system of plants. Since the earthquake time, the plant had lost external electricity, impacting the pumping working, drive by electric engine. When operable, the BWR plants responded the management of steam. However, the lack of electricity had degraded the plant maneuvers. In this paper we have presented a scheme to use the steam as an alternative drive to maintain operable the cooling system of nuclear power plant. This scheme adds more reliability and robustness to the cooling systems. Additionally, we purposed a solution to the cooling in case of lacking water for the condenser system. In our approach, steam driven turbines substitute electric engines in the ultimate emergency cooling system. (authors)

Souza Dos Santos, R. [Instituto de Engenharia Nuclear CNEN/IEN, Cidade Universitaria, Rua Helio de Almeida, 75 - Ilha do Fundiao, 21945-970 Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores / CNPq (Brazil)

2012-07-01T23:59:59.000Z

233

Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada  

SciTech Connect (OSTI)

The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

Shott, Gregory [NSTec

2014-08-31T23:59:59.000Z

234

Feasibility of breeding in hard spectrum boiling water reactors with oxide and nitride fuels  

E-Print Network [OSTI]

This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel ...

Feng, Bo, Ph. D. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

235

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

SciTech Connect (OSTI)

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

Austin, W.; Brinkley, D.

2011-10-13T23:59:59.000Z

236

Electrochemistry of Water-Cooled Nuclear Reactors  

SciTech Connect (OSTI)

This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

2006-08-08T23:59:59.000Z

237

Containment system for supercritical water oxidation reactor  

DOE Patents [OSTI]

A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

Chastagner, P.

1994-07-05T23:59:59.000Z

238

Use of phenomena identification and ranking (PIRT) process in research related to design certification of the AP600 advanced passive light water reactor (LWR)  

SciTech Connect (OSTI)

The AP600 LWR is a new advanced passive design that has been submitted to the USNRC for design certification. Within the certification process the USNRC will perform selected system thermal hydraulic response audit studies to help confirm parts of the vendor`s safety analysis submittal. Because of certain innovative design features of the safety systems, new experimental data and related advances in the system thermal hydraulic analysis computer code are being developed by the USNRC. The PIRT process is being used to focus the experimental and analytical work to obtain a sufficient and cost effective research effort. The objective of this paper is to describe the application and most significant results of the PIRT process, including several innovative features needed in the application to accommodate the short design certification schedule. The short design certification schedule has required that many aspects of the USNRC experimental and analytical research be performed in parallel, rather than in series as was normal for currently operating LWRS. This has required development and use of management techniques that focus and integrate the various diverse parts of the research. The original PIRTs were based on inexact knowledge of an evolving reactor design, and concentrated on the new passive features of the design. Subsequently, the PIRTs have evolved in two more stages as the design became more firm and experimental and analytical data became available. A fourth and final stage is planned and in progress to complete the PIRT development. The PIRTs existing at the end of each development stage have been used to guide the experimental program, scaling analyses and code development supporting the audit studies.

Wilson, G.E.; Fletcher, C.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Eltawila, F. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1996-07-01T23:59:59.000Z

239

Columbia Water and Light- Solar Rebates  

Broader source: Energy.gov [DOE]

Columbia Water and Light (CWL) offers rebates to its commercial and residential customers for the purchase of solar water heaters and solar photovoltaic systems. These rebates are available for...

240

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents [OSTI]

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Forsberg, C.W.

1985-02-19T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Development of Materials for Supercritical-Water-Cooled Reactor  

Broader source: Energy.gov [DOE]

Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system...

242

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents [OSTI]

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Forsberg, Charles W. (Kingston, TN)

1987-01-01T23:59:59.000Z

243

Advanced ceramic cladding for water reactor fuel  

SciTech Connect (OSTI)

Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

Feinroth, H.

2000-07-01T23:59:59.000Z

244

Robustness of RISMC Insights under Alternative Aleatory/Epistemic Uncertainty Classifications: Draft Report under the Risk-Informed Safety Margin Characterization (RISMC) Pathway of the DOE Light Water Reactor Sustainability Program  

SciTech Connect (OSTI)

The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, would be founded on probabilistic characterizations of uncertainty in SSC performance. In the context of probabilistic risk assessment (PRA) technology, there has arisen a general consensus about the distinctive roles of two types of uncertainty: aleatory and epistemic, where the former represents irreducible, random variability inherent in a system, whereas the latter represents a state of knowledge uncertainty on the part of the analyst about the system which is, in principle, reducible through further research. While there is often some ambiguity about how any one contributing uncertainty in an analysis should be classified, there has nevertheless emerged a broad consensus on the meanings of these uncertainty types in the PRA setting. However, while RISMC methodology shares some features with conventional PRA, it will nevertheless be a distinctive methodology set. Therefore, the paradigms for classification of uncertainty in the PRA setting may not fully port to the RISMC environment. Yet the notion of risk-informed margin is based on the characterization of uncertainty, and it is therefore critical to establish a common understanding of uncertainty in the RISMC setting.

Unwin, Stephen D.; Eslinger, Paul W.; Johnson, Kenneth I.

2012-09-20T23:59:59.000Z

245

Light Water Reactor Sustainability Program BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates  

SciTech Connect (OSTI)

As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

Sebastien Teysseyre

2014-04-01T23:59:59.000Z

246

Water inventory management in condenser pool of boiling water reactor  

DOE Patents [OSTI]

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Gluntz, D.M.

1996-03-12T23:59:59.000Z

247

Water inventory management in condenser pool of boiling water reactor  

DOE Patents [OSTI]

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Gluntz, Douglas M. (San Jose, CA)

1996-01-01T23:59:59.000Z

248

2012 Accomplishments Report | Light Water Reactor Sustainability  

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

* Completed development plan for silicon carbide (SiC) ceramic matrix composite (CMC) nuclear fuel cladding; this plan will guide future R&D on advanced cladding * Completed...

249

Candidate Materials Evaluation for Supercritical Water-Cooled Reactor  

SciTech Connect (OSTI)

Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept.

T. R. Allen and G. S. Was

2008-12-12T23:59:59.000Z

250

A coupled neutronics/thermalhydraulics tool for calculating fluctuations in Pressurized Water Reactors  

E-Print Network [OSTI]

Water Reactors or Heavy Water Reactors (LarssoA coupled neutronics/thermal­hydraulics tool for calculating fluctuations in Pressurized Water in neutron flux, fuel temperature, moderator den- sity and flow velocity in Pressurized Water Reactors

Demazière, Christophe

251

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Charges Relating to Nuclear Reactor Safety," 1976, availableissues impor tant to nuclear reactor safety. This report wasstudies of overall nuclear reactor safety have been

Nero, A.V.

2010-01-01T23:59:59.000Z

252

State space modeling of reactor core in a pressurized water reactor  

SciTech Connect (OSTI)

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10T23:59:59.000Z

253

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More DocumentsCommunicationsProvides an overview of

254

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More DocumentsCommunicationsProvides an overview

255

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More DocumentsCommunicationsProvides an overviewMilestone

256

Sustained water cleavage by visible light  

SciTech Connect (OSTI)

Sustained cleavage of water by 4 quanta of visible light is achieved in aqueous solutions by using a bifunctional redox catalyst composed of Pt and RuO/sub 2/ cosupported by colloidal TiO/sub 2/ particles. A photochemical model system containing Ru(bpy)/sub 3//sup 2 +/ as a sensitizer and methyl viologen (MV/sup 2 +/) as an electron relay is used to test the effect of catalyst composition, sensitizer concentration, pH, and temperature on the efficiency of light-induced water decomposition. Electron relay free systems also exhibit high photoactivity. Direct band gap irradiation by uv light leads to efficient water cleavage in the absence of sensitizer and relay.

Borgarello, E.; Kiwi, J.; Pelizzetti, E.; Visca, M.; Graetzel, M.

1981-10-21T23:59:59.000Z

257

Evolution of the core physics concept for the Canadian supercritical water reactor  

SciTech Connect (OSTI)

The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

2013-07-01T23:59:59.000Z

258

CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR  

SciTech Connect (OSTI)

The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

Vinson, Dennis

2010-06-01T23:59:59.000Z

259

Thermophysical properties of saturated light and heavy water for Advanced Neutron Source applications  

SciTech Connect (OSTI)

The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor`s nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300{degrees}C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250{degrees}C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

Crabtree, A.; Siman-Tov, M.

1993-05-01T23:59:59.000Z

260

Process for treating effluent from a supercritical water oxidation reactor  

DOE Patents [OSTI]

A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

Barnes, C.M.; Shapiro, C.

1997-11-25T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor  

SciTech Connect (OSTI)

Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

Mardiansah, Deby; Takaki, Naoyuki [Course of Applied Science, School of Engineering, Tokai University (Japan)

2010-06-22T23:59:59.000Z

262

Actinide minimization using pressurized water reactors  

E-Print Network [OSTI]

Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

Visosky, Mark Michael

2006-01-01T23:59:59.000Z

263

Cedarburg Light and Water Utility- Commercial Energy Efficiency Rebate Program  

Broader source: Energy.gov [DOE]

Cedarburg Light and Water Utility provides incentives for commercial, industrial and agricultural customers to increase the energy efficiency of eligible facilities. Upon request, Cedarburg Light...

264

Comparison of actinide production in traveling wave and pressurized water reactors  

SciTech Connect (OSTI)

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

Osborne, A.G.; Smith, T.A.; Deinert, M.R. [Department of Mechanical Engineering, University of Texas at Austin, Austin, TX (United States)

2013-07-01T23:59:59.000Z

265

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network [OSTI]

Nuclear Power Reactors PROTECTION AGAINST SABOTAGE Protection Against Industrial Sabotage I1C-4 Decominarion and Decommissioning

Nero, A.V.

2010-01-01T23:59:59.000Z

266

Antineutrino monitoring for the Iranian heavy water reactor  

E-Print Network [OSTI]

In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

Christensen, Eric; Jaffke, Patrick; Shea, Thomas

2014-01-01T23:59:59.000Z

267

Antineutrino monitoring for the Iranian heavy water reactor  

E-Print Network [OSTI]

In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

Eric Christensen; Patrick Huber; Patrick Jaffke; Thomas Shea

2014-03-27T23:59:59.000Z

268

Thermophysical properties of saturated light and heavy water for advanced neutron source applications  

SciTech Connect (OSTI)

The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor's nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300[degrees]C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250[degrees]C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

Crabtree, A.; Siman-Tov, M.

1993-05-01T23:59:59.000Z

269

Aging considerations for PWR (pressurized water reactor) control rod drive mechanisms and reactor internals  

SciTech Connect (OSTI)

This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors.

Ware, A.G.

1988-01-01T23:59:59.000Z

270

E-Print Network 3.0 - advanced water reactor Sample Search Results  

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

Water... it can be built on time and budget. Reactors currently under construction in Finland and France... are indeed well behind schedule. But there are several reactors that...

271

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect (OSTI)

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

272

Nuclear reactor engineering  

SciTech Connect (OSTI)

A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

Glasstone, S.; Sesonske, A.

1982-07-01T23:59:59.000Z

273

Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7  

Broader source: Energy.gov [DOE]

RELAP-7 is a nuclear reactor system safety analysis code where initial capabilities were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels.

274

A Qualitative Assessment of Thorium-Based Fuels in Supercritical Pressure Water Cooled Reactors  

SciTech Connect (OSTI)

The requirements for the next generation of reactors include better economics and safety, waste minimization (particularly of the long-lived isotopes), and better proliferation resistance (both intrinsic and extrinsic). A supercritical pressure water cooled reactor has been chosen as one of the lead contenders as a Generation IV reactor due to the high thermal efficiency and compact/simplified plant design. In addition, interest in the use of thorium-based fuels for Generation IV reactors has increased based on the abundance of thorium, and the minimization of transuranics in a neutron flux; as plutonium (and thus the minor actinides) is not a by-product in the thorium chain. In order to better understand the possibility of the combination of these concepts to meet the Generation IV goals, the qualitative burnup potential and discharge isotopics of thorium and uranium fuel were studied using pin cell analyses in a supercritical pressure water cooled reactor environment. Each of these fertile materials were used in both nitride and metallic form, with light water reactor grade plutonium and minor actinides added. While the uranium-based fuels achieved burnups that were 1.3 to 2.7 times greater than their thorium-based counterparts, the thorium-based fuels destroyed 2 to 7 times more of the plutonium and minor actinides. The fission-to-capture ratio is much higher in this reactor as compared to PWR’s and BWR’s due to the harder neutron spectrum, thus allowing more efficient destruction of the transuranic elements. However, while the uranium-based fuels do achieve a net depletion of plutonium and minor actinides, the breeding of these isotopes limits this depletion; especially as compared to the thorium-based fuels.

Weaver, Kevan Dean; Mac Donald, Philip Elsworth

2002-10-01T23:59:59.000Z

275

Gravity Scaling of a Power Reactor Water Shield  

SciTech Connect (OSTI)

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

Reid, Robert S.; Pearson, J. Boise [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

2008-01-21T23:59:59.000Z

276

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)  

SciTech Connect (OSTI)

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-04-01T23:59:59.000Z

277

austenitic light water: Topics by E-print Network  

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

First, a solid SiC-coated ... Hejzlar, P. 26 Overview of light waterhydrogen-base low energy nuclear reactions CiteSeer Summary: This paper reviews light water and hydrogen-based...

278

Water Cooling of High Power Light Emitting Diode Henrik Srensen  

E-Print Network [OSTI]

Water Cooling of High Power Light Emitting Diode Henrik Sørensen Department of Energy Technology and product lifetime. The high power Light Emitting Diodes (LED) belongs to the group of electronics

Berning, Torsten

279

COLLOQUIUM: CASL: Consortium for Advanced Simulation of Light...  

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

light water reactors (LWRs). This environment, designated the Virtual Environment for Reactor Applications (VERA), incorporates science-based models, state-of-the-art numerical...

280

Upper internals arrangement for a pressurized water reactor  

DOE Patents [OSTI]

In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

2013-07-09T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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.


281

Duquesne Light Company- Residential Solar Water Heating Program  

Broader source: Energy.gov [DOE]

Duquesne Light provides rebates to its residential customers for purchasing and installing qualifying solar water heating systems. Eligible systems may receive a flat rebate of $286 per qualifying...

282

City Water Light and Power- Residential Energy Efficiency Rebate Programs  

Broader source: Energy.gov [DOE]

City Water Light and Power (CWLP) offers rebates to Springfield residential customers for increasing the energy efficiency of participating homes. Rebates are available for geothermal heat pumps,...

283

City Water Light and Power- Commercial Energy Efficiency Rebate Programs  

Broader source: Energy.gov [DOE]

City Water Light and Power (CWLP) offers rebates to help commercial customers increase the energy efficiency of participating facilities. Energy efficient air-to-air, geothermal and water-loop...

284

Water chemistry of breeder reactor steam generators. [LMFBR  

SciTech Connect (OSTI)

The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed.

Simpson, J.L.; Robles, M.N.; Spalaris, C.N.; Moss, S.A.

1980-08-01T23:59:59.000Z

285

Advanced Water-Gas Shift Membrane Reactor  

SciTech Connect (OSTI)

The overall objectives for this project were: (1) to identify a suitable PdCu tri-metallic alloy membrane with high stability and commercially relevant hydrogen permeation in the presence of trace amounts of carbon monoxide and sulfur; and (2) to identify and synthesize a water gas shift catalyst with a high operating life that is sulfur and chlorine tolerant at low concentrations of these impurities. This work successfully achieved the first project objective to identify a suitable PdCu tri-metallic alloy membrane composition, Pd{sub 0.47}Cu{sub 0.52}G5{sub 0.01}, that was selected based on atomistic and thermodynamic modeling alone. The second objective was partially successful in that catalysts were identified and evaluated that can withstand sulfur in high concentrations and at high pressures, but a long operating life was not achieved at the end of the project. From the limited durability testing it appears that the best catalyst, Pt-Re/Ce{sub 0.333}Zr{sub 0.333}E4{sub 0.333}O{sub 2}, is unable to maintain a long operating life at space velocities of 200,000 h{sup -1}. The reasons for the low durability do not appear to be related to the high concentrations of H{sub 2}S, but rather due to the high operating pressure and the influence the pressure has on the WGS reaction at this space velocity.

Sean Emerson; Thomas Vanderspurt; Susanne Opalka; Rakesh Radhakrishnan; Rhonda Willigan

2009-01-07T23:59:59.000Z

286

TA-2 water boiler reactor decommissioning (Phase 1)  

SciTech Connect (OSTI)

Removal of external structures and underground piping associated with the gaseous effluent (stack) line from the TA-2 Water Boiler Reactor was performed as Phase I of reactor decommissioning. Six concrete structures were dismantled and 435 ft of contaminated underground piping was removed. Extensive soil contamination by /sup 137/Cs was encountered around structure TA-2-48 and in a suspected leach field near the stream flowing through Los Alamos Canyon. Efforts to remove all contaminated soil were hampered by infiltrating ground water and heavy rains. Methods, cleanup guidelines, and ALARA decisions used to successfully restore the area are described. The cost of the project was approximately $320K; 970 m/sup 3/ of low-level solid radioactive waste resulted from the cleanup operations.

Elder, J.C.; Knoell, C.L.

1986-12-01T23:59:59.000Z

287

Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability  

SciTech Connect (OSTI)

Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment. • Reactor pressure vessel • Pumps and piping

Philip E. MacDonald

2003-09-01T23:59:59.000Z

288

Corrosion Behavior of Candidate Alloys for Supercritical Water Reactors  

SciTech Connect (OSTI)

The corrosion and stress corrosion cracking behavior of metallic cladding and other core internal structures is critical to the success of the Generation IV Supercritical Water-cooled Reactors (SCWR). The eventual materials selected will be chosen based on the combined corrosion, stress-corrosion, mechanical performance, and radiation stability properties. Among the materials being considered are austenitic stainless steels, ferritic/martensitic steels, and nickel-base alloys. This paper reports initial studies on the corrosion performance of the candidate alloys 316 austenitic stainless steel, Inconel 718, and Zircaloy-2, all exposed to supercritical water at 300-500 deg. C in a corrosion loop at the University of Wisconsin. Long-term corrosion performance of AISI 347, also a candidate austenitic steel, has also been examined by sectioning samples from a component that was exposed for a period of about 30 years in supercritical water at the Genoa 3 Supercritical Water fossil power plant located in Genoa, Wisconsin. (authors)

Sridharan, K.; Zillmer, A.; Licht, J.R.; Allen, T.R.; Anderson, M.H.; Tan, L. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States)

2004-07-01T23:59:59.000Z

289

Transpiring wall supercritical water oxidation reactor salt deposition studies  

SciTech Connect (OSTI)

Sandia National Laboratories has teamed with Foster Wheeler Development Corp. and GenCorp, Aerojet to develop and evaluate a new supercritical water oxidation reactor design using a transpiring wall liner. In the design, pure water is injected through small pores in the liner wall to form a protective boundary layer that inhibits salt deposition and corrosion, effects that interfere with system performance. The concept was tested at Sandia on a laboratory-scale transpiring wall reactor that is a 1/4 scale model of a prototype plant being designed for the Army to destroy colored smoke and dye at Pine Bluff Arsenal in Arkansas. During the tests, a single-phase pressurized solution of sodium sulfate (Na{sub 2}SO{sub 4}) was heated to supercritical conditions, causing the salt to precipitate out as a fine solid. On-line diagnostics and post-test observation allowed us to characterize reactor performance at different flow and temperature conditions. Tests with and without the protective boundary layer demonstrated that wall transpiration provides significant protection against salt deposition. Confirmation tests were run with one of the dyes that will be processed in the Pine Bluff facility. The experimental techniques, results, and conclusions are discussed.

Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G. [and others

1996-09-01T23:59:59.000Z

290

Transactions of the nineteenth water reactor safety information meeting  

SciTech Connect (OSTI)

This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

Weiss, A.J. (comp.)

1991-10-01T23:59:59.000Z

291

Neurocontrol of Pressurized Water Reactors in Load-Follow Operations  

SciTech Connect (OSTI)

The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller.

Lin Chaung; Shen Chihming

2000-12-15T23:59:59.000Z

292

Irradiation behavior of pressurized water reactor control materials  

SciTech Connect (OSTI)

Postirradiation examinations have been conducted as part of an extensive Babcock and Wilcox (B and W) program in reactor control materials performance characterization. These examinations of fixed burnable poison rods and control rods confirmed operational performance and extended the material behavior data base for irradiated absorber materials used in B and W-designed pressurized water reactors. These examinations included visual, dimensional, and destructive examinations. They were conducted at B and W's Lynchburg Research Center hot cell facilities on Ag-In-Cd control rods. Al/sub 2/O/sub 3/-B/sub 4/C burnable poison rods, and B/sub 4/C control rods. The visual and dimensional exams revealed no discernible exterior damage on any of these components. Destructive examinations provided data on absorber swelling, gas release, and open porosity.

Demars, R.V.; Dideon, C.G.; Pardue, E.B.S.; Pavinich, W.A.; Thornton, T.A.; Tulenko, J.S.

1983-07-01T23:59:59.000Z

293

Pressurized water nuclear reactor system with hot leg vortex mitigator  

DOE Patents [OSTI]

A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

Lau, Louis K. S. (Monroeville, PA)

1990-01-01T23:59:59.000Z

294

Pressurized water reactor in-core nuclear fuel management by tabu search  

E-Print Network [OSTI]

Optimization of the arrangement of fuel assemblies and burnable poisons when reloading pressurized water reactors has, in the past, been performed with many di erent algorithms in an attempt to make reactors more economic and fuel effi cient...

Hill, Natasha J.; Parks, Geoffrey T.

2014-08-24T23:59:59.000Z

295

Boiling-Water Reactor internals aging degradation study. Phase 1  

SciTech Connect (OSTI)

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

Luk, K.H. [Oak Ridge National Lab., TN (United States)

1993-09-01T23:59:59.000Z

296

Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors  

SciTech Connect (OSTI)

The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction ?) for coolant voiding as standard NU fuel. (authors)

Bhatti, Zaki; Hyland, B.; Edwards, G.W.R. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01T23:59:59.000Z

297

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR (pressurized-water-reactor) plants  

SciTech Connect (OSTI)

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1988-01-01T23:59:59.000Z

298

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect (OSTI)

The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31T23:59:59.000Z

299

Camera Inspection Arm for Boiling Water Reactors - 13330  

SciTech Connect (OSTI)

Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

Martin, Scott; Rood, Marc [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)] [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)

2013-07-01T23:59:59.000Z

300

Supercritical Water Reactor Cycle for Medium Power Applications  

SciTech Connect (OSTI)

Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump and pipes were modeled with realistic assumptions using the PEACE module of Thermoflex. A three-dimensional layout of the plant was also generated with the SolidEdge software. The results of the engineering design are as follows: (i) The cycle achieves a net thermal efficiency of 24.13% with 350/460 C reactor inlet/outlet temperatures, {approx}250 bar reactor pressure and 0.75 bar condenser pressure. The steam quality at the turbine outlet is 90% and the total electric consumption of the pumps is about 2500 kWe at nominal conditions. (ii) The overall size of the plant is attractively compact and can be further reduced if a printed-circuit-heat-exchanger (vs shell-and-tube) design is used for the feedwater heater, which is currently the largest component by far. Finally, an analysis of the plant performance at off-nominal conditions has revealed good robustness of the design in handling large changes of thermal power and seawater temperature.

BD Middleton; J Buongiorno

2007-04-25T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Columbia Water and Light- Residential HVAC Rebate Program  

Broader source: Energy.gov [DOE]

Columbia Water and Light (CWL) provides an HVAC incentive for residential customers that are replacing an older heating and cooling system. Customers should submit the mechanical permit from a...

302

Columbia Water and Light- Residential Super Saver Loans  

Broader source: Energy.gov [DOE]

The Columbia Water and Light (CWL) Home Performance Super Saver Loan allows Columbia residents to finance energy improvements to homes with affordable, low interest loans with five to ten year...

303

Columbia Water and Light- Commercial Super Saver Loans  

Broader source: Energy.gov [DOE]

Columbia Water and Light (CWL) provides Commercial Super Saver Loans, which allow C&I rate customers to replace a furnace along with a new central air conditioner or heat pump with an...

304

Heat removal aspects of Liquid Metal Fast Breeder Reactor safety in light of the Three Mile Island incident  

SciTech Connect (OSTI)

The safety aspects of the Liquid Metal Fast Breeder Reactor (LMFBR) loop design are compared with those of the Light Water Reactor (LWR), in light of the Three Mile Island (TMI) incident. The events at TMI are briefly described, the fundamental differences between the LWR water coolant and the LMFBR sodium coolant are presented, and the design of analogous LMFBR safety systems under similar events as those at TMI is discussed. A preliminary qualitative evaluation of a TMI-equivalent accident for an LMFBR indicates that there is likely to be: (1) negligible pressure transients in the primary loop, (2) no core damage, (3) isolation of the incident at the steam generator, and (4) no radiation release to the environment, except a negligible amount of tritium from the secondary sodium. Furthermore, with the absence of the ECCS (Emergency Core Cooling System), pressurizer, and other pressure-related components in the LMFBR design, operator action for a LMFBR should be much simpler in dealing with the coolant upset condition and the decay heat removal problems.

Victor, H.R.; Graf, D.G.

1980-12-01T23:59:59.000Z

305

Reactor Materials Program process water piping indirect failure frequency  

SciTech Connect (OSTI)

Following completion of the probabilistic analyses, the LOCA Definition Project has been subject to various external reviews, and as a result the need for several revisions has arisen. This report updates and summarizes the indirect failure frequency analysis for the process water piping. In this report, a conservatism of the earlier analysis is removed, supporting lower failure frequency estimates. The analysis results are also reinterpreted in light of subsequent review comments.

Daugherty, W.L.

1989-10-30T23:59:59.000Z

306

ORNL/TM-2012/380 Roadmap for Nondestructive Evaluation of Reactor...  

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

by the Light Water Reactor Sustainability Program September 2012 Prepared by Cyrus Smith Randy Nanstad Robert Odette Dwight Clayton Katie Matlack Pradeep Ramuhalli Glenn Light...

307

Aging study of boiling water reactor high pressure injection systems  

SciTech Connect (OSTI)

The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1995-03-01T23:59:59.000Z

308

Nonlinear dynamics and chaos in boiling water reactors  

SciTech Connect (OSTI)

There are currently 72 commercial boiling water reactors (BWRs) in operation or under construction in the western world, 37 of them in the United States. Consequently, a great effort has been devoted to the study of BWR systems under a wide range of plant operating conditions. This paper represents a contribution to this ongoing effort; its objective is to study the basic dynamic processes in BWR systems, with special emphasis on the physical interpretation of BWR dynamics. The main thrust in this work is the development of phenomenological BWR models suited for analytical studies performed in conjunction with numerical calculations. This approach leads to a deeper understanding of BWR dynamics and facilitates the interpretation of numerical results given by currently available sophisticated BWR codes. 6 refs., 14 figs., 2 tabs.

March-Leuba, J.

1988-01-01T23:59:59.000Z

309

Pressurized water reactor fuel assembly subchannel void fraction measurement  

SciTech Connect (OSTI)

The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

Akiyama, Yoshiei [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan). Nuclear Fuel and Core Engineering Dept.; Hori, Keiichi [Mitsubishi Heavy Industries, Ltd., Hyougo (Japan); Miyazaki, Keiji [Osaka Univ. (Japan). Faculty of Engineering; Mishima, Kaichiro [Kyoto Univ., Osaka (Japan). Research Reactor Inst.; Sugiyama, Shigekazu [Nuclear Power Engineering Corp., Tokyo (Japan). Nuclear Fuel Dept.

1995-12-01T23:59:59.000Z

310

Reduced heat flow in light water (H2O) due to heavy water (D2O)  

E-Print Network [OSTI]

The flow of heat, from top to bottom, in a column of light water can be decreased by over 1000% with the addition of heavy water. A column of light water cools from 25 C to 0 C in 11 hours, however, with the addition of heavy water it takes more than 100 hours. There is a concentration dependence where the cooling time increases as the concentration of added (D2O) increases, with a near maximum being reached with as little as 2% of (D2O) added. This phenomenon will not occur if the water is mixed after the heavy water is added.

William R. Gorman; James D. Brownridge

2008-09-04T23:59:59.000Z

311

Development of Novel Water-Gas Shift Membrane Reactor  

SciTech Connect (OSTI)

This report summarizes the objectives, technical barrier, approach, and accomplishments for the development of a novel water-gas-shift (WGS) membrane reactor for hydrogen enhancement and CO reduction. We have synthesized novel CO{sub 2}-selective membranes with high CO{sub 2} permeabilities and high CO{sub 2}/H{sub 2} and CO{sub 2}/CO selectivities by incorporating amino groups in polymer networks. We have also developed a one-dimensional non-isothermal model for the countercurrent WGS membrane reactor. The modeling results have shown that H{sub 2} enhancement (>99.6% H{sub 2} for the steam reforming of methane and >54% H{sub 2} for the autothermal reforming of gasoline with air on a dry basis) via CO{sub 2} removal and CO reduction to 10 ppm or lower are achievable for synthesis gases. With this model, we have elucidated the effects of system parameters, including CO{sub 2}/H{sub 2} selectivity, CO{sub 2} permeability, sweep/feed flow rate ratio, feed temperature, sweep temperature, feed pressure, catalyst activity, and feed CO concentration, on the membrane reactor performance. Based on the modeling study using the membrane data obtained, we showed the feasibility of achieving H{sub 2} enhancement via CO{sub 2} removal, CO reduction to {le} 10 ppm, and high H{sub 2} recovery. Using the membrane synthesized, we have obtained <10 ppm CO in the H{sub 2} product in WGS membrane reactor experiments. From the experiments, we verified the model developed. In addition, we removed CO{sub 2} from a syngas containing 17% CO{sub 2} to about 30 ppm. The CO{sub 2} removal data agreed well with the model developed. The syngas with about 0.1% CO{sub 2} and 1% CO was processed to convert the carbon oxides to methane via methanation to obtain <5 ppm CO in the H{sub 2} product.

Ho, W. S. Winston

2004-12-29T23:59:59.000Z

312

Linear Parameter-Varying versus Linear Time-Invariant Control Design for a Pressurized Water Reactor  

E-Print Network [OSTI]

-dependent control to a nuclear pressurized water reactor is investigated and is compared to that of using an H1Linear Parameter-Varying versus Linear Time-Invariant Control Design for a Pressurized Water Reactor Pascale Bendotti y Electricit e de France Direction des Etudes et Recherches 6 Quai Watier, 78401

Bodenheimer, Bobby

313

Study of plutonium disposition using existing GE advanced Boiling Water Reactors  

SciTech Connect (OSTI)

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

Not Available

1994-06-01T23:59:59.000Z

314

In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies  

SciTech Connect (OSTI)

A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

2008-04-16T23:59:59.000Z

315

Light water reactor fuel response during RIA experiments  

SciTech Connect (OSTI)

Presented are a discussion of fuel rod thermal response during the RIA, a brief overview of previous test results, a discussion of the results of the PBF tests performed to date, conclusions that can be drawn from these results, and a description of the four tests remaining in the RIA testing program.

McCardell, R.K.; MacDonald, P.E.; Martinson, Z.R.; Fukuda, S.K.

1980-01-01T23:59:59.000Z

316

Light Water Reactor Sustainability Newsletter Kathryn McCarthy  

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

References * Coble, J. and J.W. Hines, 2009, "Development of a MATLAB- based Process and Equipment Prognostics Toolbox," 2009 Integrated Systems Health Management...

317

Light Water Reactor Sustainability Newsletter Kathryn A. McCarthy  

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

8 7 6 5 4 3 25 110 120 80 80 80 100 120 61 115 60 100 120 25 27 4 LWRS Newsletter Curtis Smith Risk-Informed Safety Margin Characterization Pathway Lead The RISMC Methodology and...

318

Microsoft Word - Light Water Reactor Sustainability Program Advanced...  

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

on operator attention demands and limitations on operator activities based on the current conduct of operations protocols. This report will identify opportunities to maximize...

319

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

Zircaloy); iii) hydrogen embrittlement of the cladding.cladding, even though hydrogen embrittlement has occurred.

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

320

The burnup dependence of light water reactor spent fuel oxidation  

SciTech Connect (OSTI)

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

Hanson, B.D.

1998-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

electricity generation capacity and operating efficiency of nuclear plants [Nuclear Plant Capacity Factor Nuclear Electricity Generationelectricity generation capacity and operating efficiency of nu- clear plants [

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

322

Neutron economic reactivity control system for light water reactors  

DOE Patents [OSTI]

A neutron reactivity control system for a LWBR incorporating a stationary seed-blanket core arrangement. The core arrangement includes a plurality of contiguous hexagonal shaped regions. Each region has a central and a peripheral blanket area juxapositioned an annular seed area. The blanket areas contain thoria fuel rods while the annular seed area includes seed fuel rods and movable thoria shim control rods.

Luce, Robert G. (Glenville, NY); McCoy, Daniel F. (Latham, NY); Merriman, Floyd C. (Rotterdam, NY); Gregurech, Steve (Scotia, NY)

1989-01-01T23:59:59.000Z

323

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

PWR Fuel CRUD," Proceedings of the TMS 2013 142nd Annual Meeting and Exhibition, March 3-7, 2013, San Antonio, TX, 2013. Tryggvason, G., S. Dabiri, B. Aboulhasanzadeh, J. Lu.,...

324

Light Water Reactor Sustainability Newsletter By George Griffith  

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

structure, and component (SSC) behavior will be coupled more closely to scenar- io phenomenology than is practical in today's simula- tion codes. The main output of R7 is a...

325

Strategic Plan for Light Water Reactor Research and Development  

SciTech Connect (OSTI)

The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

None

2004-02-01T23:59:59.000Z

326

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperience

327

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannel Methods for the

328

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannel Methods for

329

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannel Methods

330

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannel MethodsVirtual

331

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannel

332

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannelAbout CASL

333

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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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 JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannelAbout

334

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannelAboutAbout CASL

335

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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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 JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional AccountExperienceSubchannelAboutAbout

336

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional

337

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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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 JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation and Benchmark Study of

338

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation and Benchmark Study

339

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation and Benchmark

340

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation and BenchmarkScIEnce and

Note: This page contains sample records for the topic "light water reactor" 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

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation and BenchmarkScIEnce

342

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation and

343

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR

344

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR Media Center News

345

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR Media Center

346

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR Media CenterMedia

347

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositionalInitial Validation andPWR Media

348

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformation InExplosion Monitoring:Home| Visitors|Upcoming Events and ‹ › Science

349

Light Water Reactor Sustainability Nondestructive Evaluation for Concrete  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't Happen toLeveraging National LaboratoriesResearch and

350

Light Water Reactor Sustainability Program - Non-Destructive Evaluation  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't Happen toLeveraging National LaboratoriesResearch andR&D

351

Light Water Reactor Sustainability Program: Integrated Program Plan |  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't Happen toLeveraging National LaboratoriesResearch

352

Light Water Reactor Sustainability Program: Materials Aging and Degradation  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't Happen toLeveraging National LaboratoriesResearchTechnical

353

Materials Degradation in Light Water Reactors: Life After 60 | Department  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector General Office0-72.pdfGeorgeDoesn't32 Master EM Project Definition Rating Indexof Energy

354

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

capacity and operating efficiency of nuclear plants [31,operating efficiency of nuclear plants in the past decades.cost of the fuel Nuclear Plant Capacity Factor Nuclear

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

355

Light Water Reactor Sustainability Technical Documents | Department of  

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

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-UpHeatMulti-Dimensionalthe10IO1OP001 Letter Report:Life-CycleDutyR&D

356

Light Water Reactor Sustainability Program - Integrated Program Plan |  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |EnergyonSupport0.pdf5 OPAM SEMIANNUAL REPORT TOJaredKansas1 -

357

Light Water Reactor Sustainability Technical Documents | Department of  

Energy Savers [EERE]

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: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |EnergyonSupport0.pdf5 OPAM SEMIANNUAL REPORT TOJaredKansas1 -Energy Initiatives »

358

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National Nuclear

359

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National NuclearCOBRA-TF COBRA line

360

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National NuclearCOBRA-TF COBRA

Note: This page contains sample records for the topic "light water reactor" 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

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National NuclearCOBRA-TF COBRADakota

362

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National NuclearCOBRA-TF

363

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National NuclearCOBRA-TFFuel

364

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National NuclearCOBRA-TFFuelHydra

365

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | National

366

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPO Advanced Model for

367

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPO Advanced Model

368

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPO Advanced

369

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPO AdvancedResearch

370

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPO

371

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPOThermal Hydraulics

372

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPOThermal

373

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA (MPOThermalValidation

374

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBA

375

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals To complete its

376

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals To complete

377

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals To complete >

378

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals To complete

379

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals To

380

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals ToAbout CASL The

Note: This page contains sample records for the topic "light water reactor" 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

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals ToAbout CASL

382

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals ToAbout CASLMedia

383

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals ToAbout

384

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals ToAboutNews

385

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoals

386

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoalsEducation Webinars

387

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoalsEducation

388

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony | NationalMAMBAGoalsEducationPartners

389

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |

390

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho National Laboratory Idaho Falls,

391

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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

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392

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho National Laboratory

393

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho National LaboratoryPerformance

394

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho National

395

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho NationalResources CASL Resources

396

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho NationalResources CASL

397

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho NationalResources CASLSoftware

398

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho NationalResources

399

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho NationalResourcesStrategy The

400

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho NationalResourcesStrategy

Note: This page contains sample records for the topic "light water reactor" 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

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |Idaho

402

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |IdahoVision Predict, with confidence,

403

Consortium for Advanced Simulation of Light Water Reactors (CASL)  

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

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404

Consortium for Advanced Simulation of Light Water Reactors (CASL) -  

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 May Jun Jul(Summary)morphinanInformation InInformationCenterResearchCASLNanoporousTestimony |IdahoVision Predict, withOne-Roof

405

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network [OSTI]

of hydride fueled BWRs. Nuclear Engineering and Design, 239:Fueled PWR Cores. Nuclear Engineering and Design, 239:1489–Hydride Fueled LWRs. Nuclear Engineering and Design, 239:

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

406

Corrosion optimized Zircaloy for boiling water reactor (BWR) fuel elements  

SciTech Connect (OSTI)

A corrosion optimized Zircaloy has to be based primarily on in-boiling water reactor (in-BWR) results. Therefore, the material parameters affecting corrosion were deduced from results of experimental fuel rod irradiation with systematic variations and from a large variety of material coupons exposed in water rods up to four cycles. The major material effects is the size and distribution of precipitates. For optimizing both early and late corrosion, the size has to stay in a small range. In the case of material quenched in the final stage, the quenching rate appears to be an important parameter. As far as materials chemistry is concerned, the in-BWR results indicate that corrosion in BWRs is influenced by the alloying elements tin, chromium, and the impurity silicon. In addition to corrosion optimization, hydriding is also considered. A large variation from lot to lot under identically coolant condition has been found. The available data indicate that the chromium content is the most important material parameter for hydrogen pickup.

Garzarolli, F.; Schumann, R.; Steinberg, E. [Siemens AG, Erlangen (Germany). Power Generation Group

1994-12-31T23:59:59.000Z

407

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

408

Evaluation of Buildup of Activated Corrosion Products for Highly Compact Marine Reactor DRX without Primary Coolant Water Purification System  

E-Print Network [OSTI]

Evaluation of Buildup of Activated Corrosion Products for Highly Compact Marine Reactor DRX without Primary Coolant Water Purification System

Odano, N

2000-01-01T23:59:59.000Z

409

Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.  

SciTech Connect (OSTI)

The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluate the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.

Yun, D.; Taiwo, T. A.; Kim, T. K.; Mohamed, A.; Nuclear Engineering Division

2010-10-01T23:59:59.000Z

410

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents [OSTI]

A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

Forsberg, Charles W. (Oak Ridge, TN)

1991-01-01T23:59:59.000Z

411

Seismicity and seismic response of the Soviet-designed VVER (Water-cooled, Water moderated Energy Reactor) reactor plants  

SciTech Connect (OSTI)

On March 4, 1977, a strong earthquake occurred at Vrancea, Romania, about 350 km from the Kozloduy plant in Bulgaria. Subsequent to this event, construction of the unit 2 of the Armenia plant was delayed over two years while seismic features were added. On December 7, 1988, another strong earthquake struck northwest Armenia about 90 km north of the Armenia plant. Extensive damage of residential and industrial facilities occurred in the vicinity of the epicenter. The earthquake did not damage the Armenia plant. Following this event, the Soviet government announced that the plant would be shutdown permanently by March 18, 1989, and the station converted to a fossil-fired plant. This paper presents the results of the seismic analyses of the Soviet-designed VVER (Water-cooled, Water moderated Energy Reactor) plants. Also presented is the information concerning seismicity in the regions where VVERs are located and information on seismic design of VVERs. The reference units are the VVER-440 model V230 (similar to the two units of the Armenia plant) and the VVER-1000 model V320 units at Kozloduy in Bulgaria. This document provides an initial basis for understanding the seismicity and seismic response of VVERs under seismic events. 1 ref., 9 figs., 3 tabs.

Ma, D.C.; Gvildys, J.; Wang, C.Y.; Spencer, B.W.; Sienicki, J.J.; Seidensticker, R.W.; Purvis, E.E. III

1989-01-01T23:59:59.000Z

412

Near-infrared light scattering by particles in coastal waters  

E-Print Network [OSTI]

Near-infrared light scattering by particles in coastal waters David Doxaran* , Marcel Babin extend over the near-infrared spectral region to up to 870 nm. The measurements were conducted in three in the near-infrared very closely matched a - spectral dependence, which is expected when the particle size

Babin, Marcel

413

Experimental Studies of NGNP Reactor Cavity Cooling System With Water  

SciTech Connect (OSTI)

This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

Michael Corradini; Mark Anderson; Yassin Hassan; Akira Tokuhiro

2013-01-16T23:59:59.000Z

414

Multi-cycle boiling water reactor fuel cycle optimization  

SciTech Connect (OSTI)

In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

Ottinger, K.; Maldonado, G.I. [University of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

2013-07-01T23:59:59.000Z

415

Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980  

SciTech Connect (OSTI)

The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

Romano, A.J. (comp.)

1980-06-01T23:59:59.000Z

416

Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980  

SciTech Connect (OSTI)

The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

1980-08-01T23:59:59.000Z

417

Water Reactor Safety Research Division. Quarterly progress report, October 1-December 31, 1980  

SciTech Connect (OSTI)

The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: Stress Corrosion Cracking of PWR Steam Generator Tubing, Advanced Code Evaluation, Simulator Improvement Program, and LWR Assessment and Application.

Cerbone, R.J.; Saha, P.; van Rooyen, D.

1981-02-01T23:59:59.000Z

418

Conceptual design of an annular-fueled superheat boiling water reactor  

E-Print Network [OSTI]

The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

419

Stability analysis of the boiling water reactor : methods and advanced designs  

E-Print Network [OSTI]

Density Wave Oscillations (DWOs) are known to be possible when a coolant undergoes considerable density reduction while passing through a heated channel. In the development of boiling water reactors (BWRs), there has been ...

Hu, Rui, Ph. D. Massachusetts Institute of Technology

2010-01-01T23:59:59.000Z

420

Design strategies for optimizing high burnup fuel in pressurized water reactors  

E-Print Network [OSTI]

This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

Xu, Zhiwen, 1975-

2003-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems  

DOE Patents [OSTI]

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

1994-05-03T23:59:59.000Z

422

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems  

DOE Patents [OSTI]

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

McDermott, Daniel J. (Export, PA); Schrader, Kenneth J. (Penn Hills, PA); Schulz, Terry L. (Murrysville Boro, PA)

1994-01-01T23:59:59.000Z

423

Analysis of strategies for improving uranium utilization in pressurized water reactors  

E-Print Network [OSTI]

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

Sefcik, Joseph A.

1981-01-01T23:59:59.000Z

424

An inverted pressurized water reactor design with twisted-tape swirl promoters  

E-Print Network [OSTI]

An Inverted Fuel Pressurized Water Reactor (IPWR) concept was previously investigated and developed by Paolo Ferroni at MIT with the effort to improve the power density and capacity of current PWRs by modifying the core ...

Nguyen, Nghia T. (Nghia Tat)

2014-01-01T23:59:59.000Z

425

The selective use of thorium and heterogeneity in uranium-efficient pressurized water reactors  

E-Print Network [OSTI]

Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the ...

Kamal, Altamash

1982-01-01T23:59:59.000Z

426

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)  

SciTech Connect (OSTI)

The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

Shaver, Mark W.; Lanning, Donald D.

2010-02-01T23:59:59.000Z

427

Radiological Control of Water in Reactor Pond of MR Reactor in NRC 'Kurchatov Institute', During Dismantling Work - 13462  

SciTech Connect (OSTI)

The analysis of the activity and radionuclide composition of water from the MR reactor pond for ?,?,?-ray radionuclides was made. To solve this problem we use a wide range of laboratory equipment: gamma spectrometric complex, beta spectrometric complex, vacuum alpha spectrometer, and spectrometric complex with liquid scintillator. The water from MR reactor pond contains: Cs-137 (2,6*10{sup 2} Bq/g), Co-60(1,8 Bq/g), Sr-90 (1,0*10{sup 2} Bq/g), H-3 (7,0*10{sup 3} Bq/g), and components of nuclear fuel (U-232,U-234,U-235,U-236,U-238). Therefore the cleaning water from radioactivity waste occurs to be quite a complicated radiochemical task. (authors)

Stepanov, Alexey; Simirsky, Yury; Semin, Ilya; Volkovich, Anatoly; Ivanov, Oleg [National Research Center 'Kurchatov Institute', Moscow (Russian Federation)] [National Research Center 'Kurchatov Institute', Moscow (Russian Federation)

2013-07-01T23:59:59.000Z

428

An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors  

SciTech Connect (OSTI)

An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

Gentry, Cole A [ORNL] [ORNL; Godfrey, Andrew T [ORNL] [ORNL; Terrani, Kurt A [ORNL] [ORNL; Gehin, Jess C [ORNL] [ORNL; Powers, Jeffrey J [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL

2014-01-01T23:59:59.000Z

429

Post-remedial action report for the Water Boiler Reactor Site  

SciTech Connect (OSTI)

The TA-2 Water Boiler Reactor Decommissioning Project decontaminated and decommissioned the Water Boiler Reactor, TA-2-1-122, at Los Alamos National Laboratory in Los Alamos, New Mexico, to provide reusable space at the TA-2 site and to eliminate the hazard of accidental intrusion into a contaminated structure. This report documents the radiological condition of the site after the decommissioning and decontamination between June 1989 and April 1990. 7 refs., 3 figs., 5 tabs.

Montoya, G.M.

1991-05-01T23:59:59.000Z

430

Advanced reactor safety research, quarterly report, October-December 1980  

SciTech Connect (OSTI)

Information is presented concerning advanced reactor core phenomenology; light water reactor severe core damage phenomenology; core debris behavior; containment analysis; elevated temperature design assessment; LMFBR accident delineation; and test and facility technology.

Not Available

1982-01-01T23:59:59.000Z

431

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)  

SciTech Connect (OSTI)

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

NONE

1994-04-30T23:59:59.000Z

432

Cyclic Mode of Transmutation of Minor Actinides in Heavy-Water Reactor  

SciTech Connect (OSTI)

Characteristics of process of transmutation of americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, {sup 244}Cm and {sup 238}Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of a transmutation, the radiotoxicity is increased by 11 times in comparison with initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute americium and curium extracted from 7-8 VVER-1000 type reactors. It means that the required power of transmutation reactor makes about 4 % of thermal power of VVER-1000 type reactors. (authors)

Gerasimov, Aleksander S.; Kiselev, Gennady V.; Myrtsymova, Lidia A.; Zaritskaya, Tamara S. [Institute of Theoretical and Experimental Physics, SSC RF ITEP, Bolshaya Cheremushkinskaya, 25, 117218 Moscow (Russian Federation)

2002-07-01T23:59:59.000Z

433

Innovative fuel designs for high power density pressurized water reactor  

E-Print Network [OSTI]

One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of ...

Feng, Dandong, Ph. D. Massachusetts Institute of Technology

2006-01-01T23:59:59.000Z

434

Calculation of heat capacities of light and heavy water by path-integral molecular dynamics  

E-Print Network [OSTI]

Calculation of heat capacities of light and heavy water by path-integral molecular dynamics-integral molecular dynamics has been used to calculate the constant-volume heat capacities of light and heavy water

Nielsen, Steven O.

435

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents [OSTI]

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Corletti, Michael M. (New Kensington, PA); Schulz, Terry L. (Murrysville, PA)

1993-01-01T23:59:59.000Z

436

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents [OSTI]

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

Corletti, M.M.; Schulz, T.L.

1993-12-07T23:59:59.000Z

437

Comparison of thorium-based fuels with different fissile components in existing boiling water reactors  

E-Print Network [OSTI]

Comparison of thorium-based fuels with different fissile components in existing boiling water, SE-412 96 Göteborg, Sweden Keywords: Thorium BWR Neutronics a b s t r a c t With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium

Demazière, Christophe

438

Vibrational spectra of light and heavy water with application to neutron cross section calculations  

SciTech Connect (OSTI)

The design of nuclear reactors and neutron moderators require a good representation of the interaction of low energy (E < 1 eV) neutrons with hydrogen and deuterium containing materials. These models are based on the dynamics of the material, represented by its vibrational spectrum. In this paper, we show calculations of the frequency spectrum for light and heavy water at room temperature using two flexible point charge potentials: SPC-MPG and TIP4P/2005f. The results are compared with experimental measurements, with emphasis on inelastic neutron scattering data. Finally, the resulting spectra are applied to calculation of neutron scattering cross sections for these materials, which were found to be a significant improvement over library data.

Damian, J. I. Marquez; Granada, J. R. [Neutron Physics Department and Instituto Balseiro, Centro Atomico Bariloche, CNEA (Argentina); Malaspina, D. C. [Department of Biomedical Engineering and Chemistry of Life Processes Institute, Northwestern University, 2145 Sheridan Road, Evanston, Illinois 60208 (United States)

2013-07-14T23:59:59.000Z

439

Accuracy Based Generation of Thermodynamic Properties for Light Water in RELAP5-3D  

SciTech Connect (OSTI)

RELAP5-3D interpolates to obtain thermodynamic properties for use in its internal calculations. The accuracy of the interpolation was determined for the original steam tables currently used by the code. This accuracy evaluation showed that the original steam tables are generally detailed enough to allow reasonably accurate interpolations in most areas needed for typical analyses of nuclear reactors cooled by light water. However, there were some regions in which the original steam tables were judged to not provide acceptable accurate results. Revised steam tables were created that used a finer thermodynamic mesh between 4 and 21 MPa and 530 and 640 K. The revised steam tables solved most of the problems observed with the original steam tables. The accuracies of the original and revised steam tables were compared throughout the thermodynamic grid.

Cliff B. Davis

2010-09-01T23:59:59.000Z

440

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative  

SciTech Connect (OSTI)

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

1996-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "light water reactor" 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.


441

Method of burning lightly loaded coal-water slurries  

DOE Patents [OSTI]

In a preferred arrangement of the method of the invention, a lightly loaded coal-water slurry, containing in the range of approximately 40% to 52% + 2% by weight coal, is atomized to strip water from coal particles in the mixture. Primary combustor air is forced around the atomized spray in a combustion chamber of a combustor to swirl the air in a helical path through the combustion chamber. A flame is established within the combustion chamber to ignite the stripped coal particles, and flame temperature regulating means are provided for maintaining the flame temperature within a desired predetermined range of temperatures that is effective to produce dry, essentially slag-free ash from the combustion process.

Krishna, C.R.

1984-07-27T23:59:59.000Z

442

Reduced heat flow in light water (H2O) due to heavy water (D2O) William R. Gormana)  

E-Print Network [OSTI]

Reduced heat flow in light water (H2O) due to heavy water (D2O) William R. Gormana) and James D by over 1000% with the addition of heavy water. A column of light water cools from 25°C to 0°C in 11 hours, however, with the addition of heavy water it takes more than 100 hours. There is a concentration

Suzuki, Masatsugu

443

Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor  

SciTech Connect (OSTI)

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Pearson, J. Boise; Stewart, Eric T. [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States); Reid, Robert S. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States)

2007-01-30T23:59:59.000Z

444

Component failures at pressurized water reactors. Final report  

SciTech Connect (OSTI)

Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

Reisinger, M.F.

1980-10-01T23:59:59.000Z

445

Wastewater Effluent Polishing Systems of Anaerobic Baffled Reactor Treating Black-water from Households  

E-Print Network [OSTI]

Wastewater Effluent Polishing Systems of Anaerobic Baffled Reactor Treating Black-water from of different integrated low-cost wastewater treatment systems, comprising one ABR as first treatment step filter and a vertical flow constructed wetland. A mixture of septage and domestic wastewater was used

Richner, Heinz

446

Water cooled metal optics for the Advanced Light Source  

SciTech Connect (OSTI)

The program for providing water cooled metal optics for the Advanced Light Source at Berkeley is reviewed with respect to fabrication and metrology of the surfaces. Materials choices, surface figure and smoothness specifications, and metrology systems for measuring the plated metal surfaces are discussed. Results from prototype mirrors and grating blanks will be presented, which show exceptionally low microroughness and mid-period error. We will briefly describe out improved version of the Long Trace Profiler, and its importance to out metrology program. We have completely redesigned the mechanical, optical and computational parts of the profiler system with the cooperation of Peter Takacs of Brookhaven, Continental Optical, and Baker Manufacturing. Most important is that one of our profilers is in use at the vendor to allow testing during fabrication. Metrology from the first water cooled mirror for an ALS beamline is presented as an example. The preplating processing and grinding and polishing were done by Tucson Optical. We will show significantly better surface microroughness on electroless nickel, over large areas, than has been reported previously.

McKinney, W.R.; Irick, S.C. [Lawrence Berkeley Lab., CA (United States); Lunt, D.L.J. [Tucson Optical Research Corp., AZ (United States)

1991-10-28T23:59:59.000Z

447

Feasibility of Water Cooled Thorium Breeder Reactor Based on LWR Technology  

SciTech Connect (OSTI)

The feasibility of Th-{sup 233}U fueled, homogenous breeder reactor based on matured conventional LWR technology was studied. The famous demonstration at Shipping-port showed that the Th-{sup 233}U fueled, heterogeneous PWR with four different lattice fuels was possible to breed fissile but its low averaged burn-up including blanket fuel and the complicated core configuration were not suitable for economically competitive reactor. The authors investigated the wide design range in terms of fuel cell design, power density, averaged discharge burn-up, etc. to determine the potential of water-cooled Th reactor as a competitive breeder. It is found that a low moderated (MFR=0.3) H{sub 2}O-cooled reactor with comparable burn-up with current LWR is feasible to breed fissile fuel but the core size is too large to be economical because of the low pellet power density. On the other hand, D{sub 2}O-cooled reactor shows relatively wider feasible design window, therefore it is possible to design a core having better neutronic and economic performance than H{sub 2}O-cooled. Both coolant-type cores show negative void reactivity coefficient while achieving breeding capability which is a distinguished characteristics of thorium based fuel breeder reactor. (authors)

Takaki, Naoyuki; Permana, Sidik; Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

2007-07-01T23:59:59.000Z

448

Lansing Board of Water and Light- Hometown Energy Savers Commercial Rebates  

Broader source: Energy.gov [DOE]

Franklin Energy Services and the Lansing Board of Water and Light (LBWL) partner together to offer the Hometown Energy Savers® Commercial and Industrial Energy Efficiency Rebate Program. Eligible...

449

McMinnville Water and Light- Residential Energy Efficiency Rebate Program  

Broader source: Energy.gov [DOE]

McMinnville Water and Light (MWL) offers rebates on energy efficient homes, appliances and equipment to their residential customers. Rebates are valid on refrigerators, freezers, clothes washer,...

450

Development of hafnium and comparison with other pressurized water reactor control rod materials  

SciTech Connect (OSTI)

Development of a special application of hafnium for pressurized water reactor control rods is discussed. A unique feature of the design is the sealing of the hafnium material inside protective stainless steel tubing, whereas in prior applications the hafnium material was exposed directly to the reactor primary coolant. A comparison is made of the new hafnium design with silver-indium-cadmium and B/sub 4/C hybrid control rod material design applications. The advantages and disadvantages of the alternative designs are summarized, including performance and fabrication considerations.

Keller, H.W.; Hollein, D.A.; Hott, A.C.; Shallenberger, J.M.

1982-12-01T23:59:59.000Z

451

In-reactor oxidation of zircaloy-4 under low water vapor pressures  

SciTech Connect (OSTI)

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Walter G. Luscher; David J. Senor; Keven K. Clayton; Glen R. Longhurst

2015-01-01T23:59:59.000Z

452

Implications for accident management of adding water to a degrading reactor core  

SciTech Connect (OSTI)

This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-02-01T23:59:59.000Z

453

Continuous fiber ceramic composite cladding for commercial water reactor fuel; Final Project  

SciTech Connect (OSTI)

This project is a research effort to develop and demonstrate the feasibility of an improved ceramics-based cladding material for water reactor fuel, which will be significantly more resistant to structural damage during a LOCA accident than the current Zircaloy cladding material. Specifically, the goal of this NERI project is to determine, via engineering type tests, the feasibility of substituting such advanced ceramic materials for the Zircaloy cladding now in use. This report presents the project research and development activities, which included prototype material design, fabrication, characterization, LOCA type of thermal shock testing, and in-reactor irradiation/corrosion testing. The report also presents the technical finding and discussions of results. The technical task were performed in collaboration with four subcontractors: The Advanced Materials Section of McDermott Technology Incorporated (MTI), the Nuclear Reactor Laboratory of Massachusetts Institute of Technology (MTI), Swales Aerospace Inc., and the Thin Film Laboratory of Northwestern University.

Herbert Feinroth

2001-04-30T23:59:59.000Z

454

Knowledge base expert system control of spatial xenon oscillations in pressurized water reactors  

SciTech Connect (OSTI)

Nuclear reactor operators are required to pay special attention to spatial xenon oscillations during the load-follow operation of pressurized water reactors. They are expected to observe the axial offset of the core, and to estimate the correct time and amount of necessary control action based on heuristic rules given in axial xenon oscillations are knowledge intensive, and heuristic in na