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1

Secretary Chu Statement on AP1000 Reactor Design Certification | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Secretary Chu Statement on AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification December 22, 2011 - 3:25pm Addthis Washington, D.C. - U.S. Energy Secretary Steven Chu issued the following statement today in support of the Nuclear Regulatory Commission's (NRC) decision to certify Westinghouse Electric's AP1000 nuclear reactor design, a significant step towards constructing a new generation of U.S. nuclear reactors. In February 2010, the Obama Administration announced the offer of a conditional commitment for a $8.33 billion loan guarantee for the construction and operation of two AP1000 reactors at Alvin W. Vogtle Electric Generation Plant in Burke, Georgia. "The Administration and the Energy Department are committed to restarting

2

Structural Design Challenges in Design Certification Applications for New Reactors  

SciTech Connect

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design chal- lenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

Miranda, M.; Braverman, J.; Wei, X.; Hofmayer, C.; Xu, J.

2011-07-17T23:59:59.000Z

3

Seismic Analysis Issues in Design Certification Applications for New Reactors  

SciTech Connect

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

Miranda, M.; Morante, R.; Xu, J.

2011-07-17T23:59:59.000Z

4

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices  

Science Conference Proceedings (OSTI)

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

Not Available

1994-07-01T23:59:59.000Z

5

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report  

SciTech Connect

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

Not Available

1994-07-01T23:59:59.000Z

6

Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1  

SciTech Connect

This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

NONE

1997-05-01T23:59:59.000Z

7

Advanced light water reactor plants system 80+{trademark} design certification program. Annual progress report, October 1, 1993--September 30, 1994  

SciTech Connect

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80{sup +}{trademark} during the U.S. government`s 1994 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems. 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 and the System 80+ design form the basis of the Korean standardization program. The Nuclear Island portion of the System 80+ standard design has also been offered to the Republic of China, in response to their bid specification for an ALWR. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was docketed by the Nuclear Regulatory Commission (NRC) in May 1991 and a Draft Safety Evaluation Report (DSER) was issued in October 1992.

Not Available

1995-01-01T23:59:59.000Z

8

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

SciTech Connect

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

9

Certification plan for reactor analysis computer codes  

Science Conference Proceedings (OSTI)

A certification plan for reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations has been prepared. An action matrix, checklists, a time schedule, and a resource commitment table have been included in the plan. These items identify what is required to achieve certification of the codes, the time table that this will be accomplished on, and the resources needed to support such an effort.

Toffer, H.; Crowe, R.D.; Schwinkendorf, K.N. [Westinghouse Hanford Co., Richland, WA (United States); Pevey, R.E. [Westinghouse Savannah River Co., Aiken, SC (United States)

1990-01-01T23:59:59.000Z

10

System 80+{trademark} Standard Design: CESSAR design certification. Volume 4: Amendment I  

Science Conference Proceedings (OSTI)

This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 4, provides a description of the reactor, reactor internals, fuel assemblies, and associated design requirements.

Not Available

1990-12-21T23:59:59.000Z

11

Remedial Action Certification Docket - Sodium Reactor Experiment (SRE)  

Office of Legacy Management (LM)

c~-?i-- c~-?i-- I ,3-l Remedial Action Certification Docket - Sodium Reactor Experiment (SRE) .Complex and the Hot Cave Facility (Bldg. 003), Santa Susana ,Fie!d Laboratory, Chatsworth, California ..:'..~::Yerlette Gatl in, MA-232 I am attaching for entry into the Public Document Room, one copy of the N -23 subject documentat ion. These documents are the backup data for the certification that the facilfties are radiologically acceptable for b- unrestricted use as noted in the certification statement published in the &aney Federal Register. Inasmuch as the certification for unrestricted use is 9/2(/85 being published in the Federal Register, it is prudent that the attached documentation also be available to the public. These documents should be retained In accordance with DOE Order 1324.2--disposal schedule 25.

12

Slurry reactor design studies  

SciTech Connect

The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

13

Web Design and Web Development Certification Program CORE COURSES FOR BOTH CERTIFICATES (exams required for completion)  

E-Print Network (OSTI)

Web Design and Web Development Certification Program CORE COURSES FOR BOTH CERTIFICATES (exams) (1-day) Introduction to Web Design & Development: Workflow That Works (4-day) HTML & CSS (2-day completion of exam, PS2 can be substituted) (3-day) *Web Graphics & Interface Design Using Photo

Duong, Timothy Q.

14

TRUPACT-2 design, testing and certification  

Science Conference Proceedings (OSTI)

A large volume of contact-handled transuranic (CH TRU) nuclear waste has been generated and has accumulated at ten Department of Energy (DOE) facilities. Approximately 168,000 cubic meters of this waste has been targeted for shipment to the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. Shipment of all of this waste is expected to require more than 15,000 shipments, totalling approximately 38 million kilometers of travel. The TRUPACT-II is a reusable package for shipment of CH TRU waste; it is designed in accordance with US Nuclear Regulatory Commission (NRC) requirements for Type B packages found in 10 CFR 71. There are two separate levels of containment to permit shipment of plutonium in excess of 740 GBq per package. The packaging is a right circular cylinder in shape, with a domed top and a flat bottom; external dimensions are 240 centimeters in diameter and 309 centimeters high. The capacity of each TRUPACT-II is 3182 kilograms of waste that can be loaded into fourteen 210-1 drums; or two 1.9 cubic meters standard waste boxes (SWBs). Three TRUPACT-II packages are transported on a custom designed semitrailer which is pulled by a conventional tractor for highway transport of CH TRU waste between DOE sites and to the WIPP near Carlsbad, New Mexico. The Safety Analysis Report (SAR) for the TRUPACT-II Package was initially submitted to the NRC in March 1989. Four revisions were issued during the period from May through August 1989. NRC certification of the package was received on August 30, 1989. 5 refs., 4 figs., 1 tab.

Cruse, J. (Pacific Nuclear Systems, Inc., Federal Way, WA (USA))

1990-01-01T23:59:59.000Z

15

Designing Reactors to Facilitate Decommissioning  

SciTech Connect

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

16

Designing decommissioning into new reactor designs  

SciTech Connect

One of the lessons learned from decommissioning of existing reactors has been that decommissioning was not given much thought when these reactors were designed some three or four decades ago. Recently, the nuclear power has seen a worldwide resurgence and many new advanced reactor designs are either on the market or nearing design completion. Most of these designs are evolutionary in nature and build on the existing and proven technologies. They also incorporate many improvements and take advantage of the substantial operating experience. Nevertheless, by and large, the main factors driving the design of new reactors are the safety features, safeguards considerations, and the economic factors. With a large decommissioning experience that already exists in the nuclear industry, and with average decommissioning costs at around six hundred million dollars for each reactor in today's dollars, it is necessary that decommissioning factors also be considered as a part of the early design effort. Even though decommissioning may be sixty years down the road from the time they go on line, it is only prudent that new designs be optimized for eventual decommissioning, along with the other major considerations. (authors)

Devgun, J.S.; CHMM, Ph.D. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States)

2007-07-01T23:59:59.000Z

17

Fuel Reformation: Microchannel Reactor Design  

DOE Green Energy (OSTI)

Fuel processing is used to extract hydrogen from conventional vehicle fuel and allow fuel cell powered vehicles to use the existing petroleum fuel infrastructure. Kilowatt scale micro-channel steam reforming, water-gas shift and preferential oxida-tion reactors have been developed capable of achieving DOE required system performance metrics. Use of a microchannel design effectively supplies heat to the highly endothermic steam reforming reactor to maintain high conversions, controls the temperature profile for the exothermic water gas shift reactor, which optimizes the overall reaction conversion, and removes heat to prevent the unwanted hydrogen oxidation in the prefer-ential oxidation reactor. The reactors combined with micro-channel heat exchangers, when scaled to a full sized 50 kWe automotive system, will be less than 21 L in volume and 52 kg in weight.

Brooks, Kriston P.; Davis, James M.; Fischer, Christopher M.; King, David L.; Pederson, Larry R.; Rawlings, Gregg C.; Stenkamp, Victoria S.; TeGrotenhuis, Ward E.; Wegeng, Robert S.; Whyatt, Greg A.

2005-09-01T23:59:59.000Z

18

Advanced burner test reactor preconceptual design report.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

2008-12-16T23:59:59.000Z

19

System 80+{trademark} Standard Design: CESSAR design certification. Volume 2: Amendment I  

SciTech Connect

This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 2, in conjunction with Volume 3, provides the design of structures, components, equipment and systems.

Not Available

1990-12-21T23:59:59.000Z

20

System 80+{trademark} Standard Design: CESSAR design certification. Volume 3: Amendment I  

SciTech Connect

This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 3, in conjunction with Volume 2, provides the design of structures, components, equipment and systems.

Not Available

1990-12-21T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Design guide for category V reactors transient reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.

Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A

1979-03-01T23:59:59.000Z

22

ESBWR... An Evolutionary Reactor Design  

Science Conference Proceedings (OSTI)

GE's latest evolution of the Boiling Water Reactor, the ESBWR, combines improvements in safety with design simplification and component standardization to produce a safer, more reliable nuclear power plant, with lower projected construction costs than plants in operation today. The ESBWR program started in the early 1990's when GE was developing the Simplified Boiling Water Reactor (SBWR). GE stopped this program because the power output of the SBWR was too small to generate the right economics for a new build project. The program was a success however, because the design proved many of the passive safety technology developments that are being utilized in the ESBWR. By harnessing these design concepts and testing results from the original SBWR and construction and operating experience from the Advanced Boiling Water Reactor (ABWR), the ESBWR design team has produced a simplified reactor with a standardized design and first-rate economics. Significant simplification of plant systems is achieved in the ESBWR. As a result, operating and maintenance staff requirements are reduced; low-level waste generation is reduced; dose rates are reduced; operational reliability is improved; and plant safety and security are improved. Each of these improvements provide distinct and unique advantages to the ESBWR design. First, fewer active components (in particular, active safety systems) reduce the maintenance and online surveillance requirements, thereby reducing operational exposure and dose rates. Second, fewer demands on plant operators and safety systems reduce plant operating staff while still providing direct improvements in accident and transient response. Finally, reductions in building volumes and required manufactured components shorten the length of time needed for ESBWR construction, resulting in improved financial returns for plant owners. The ESBWR is designed to meet the needs of nuclear power plant owners today and into the future, with a 60-year design life. Through design simplification and standardization, ESBWR offers improved safety, increased reliability, and ease of operation. Yet compared to current nuclear power plants, the ESBWR requires only a fraction of traditional plant operating and maintenance staff, offers faster construction and lower costs of construction, while also reducing operational costs. (authors)

Gamble, Robert E. [GE Energy - Nuclear, 1989 Little Orchard St., San Jose, CA 95125 (United States); Hinds, David H.; Hucik, Steven A.; Maslak, Chris E. [GE Energy - Nuclear, 3901 Castle Hayne Road A-30, Wilmington, NC 28402 (United States)

2006-07-01T23:59:59.000Z

23

System 80+{trademark} Standard Design: CESSAR design certification. Volume 9: Amendment I  

Science Conference Proceedings (OSTI)

This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 9 discusses Electric Power and Auxiliary Systems.

Not Available

1990-12-21T23:59:59.000Z

24

System 80+{trademark} Standard Design: CESSAR design certification. Volume 8: Amendment I  

SciTech Connect

This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 8 provides a description of instrumentation and controls.

Not Available

1990-12-21T23:59:59.000Z

25

System 80+{trademark} Standard Design: CESSAR design certification. Volume 18: Amendment I  

SciTech Connect

This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80+{trademark} Standard Design. This Volume 18 provides Appendix B, Probabilistic Risk Assessment.

Not Available

1990-12-21T23:59:59.000Z

26

System 80+{trademark} Standard Design: CESSAR design certification. Volume 11: Amendment I  

Science Conference Proceedings (OSTI)

This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 11 discusses Radiation Protection, Conduct of Operations, and the Initial Test Program.

Not Available

1990-12-21T23:59:59.000Z

27

System 80+{trademark} Standard Design: CESSAR design certification. Volume 10: Amendment I  

Science Conference Proceedings (OSTI)

This report, entitled Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These volumes describe the Combustion Engineering, Inc. System 80{sup +}{trademark} Standard Design. This volume 10 discusses the Steam and Power Conversion System and Radioactive Waste Management.

Not Available

1990-12-21T23:59:59.000Z

28

Russian RBMK reactor design information  

Science Conference Proceedings (OSTI)

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

Not Available

1993-11-01T23:59:59.000Z

29

PRISM; The plant design concept for the U. S. advanced liquid metal reactor program  

SciTech Connect

The US program for development of an advanced liquid metal reactor (ALMR) is proceeding into a new phase of focused design development. This new phase started at the beginning of 1989; its objective is to complete the conceptual design of the US ALMR, with supporting key feature tests, sufficiently to enter a more detailed design phase and subsequent construction of a prototype reactor plant. A project goal is to demonstrate by actual performance of the reactor its passive, inherent safety features and thereby provide the technical basis for certification of the design by the Nuclear Regulatory Commission (NRC). This paper reports on the PRISM (power reactor inherently safe module) reactor concept which in combination with the IFR (integral fast reactor) metal fuel cycle being developed by Argonne National Laboratory, was selected by DOE in 1988 as the reference design for the US ALMR program.

Berglund, R.C.; Tippets, F.E. (GE Nuclear Energy, Advance Nuclear Technology, San Jose, CA (US))

1989-01-01T23:59:59.000Z

30

Reactor closure design for a pool-type fast reactor  

SciTech Connect

The reactor closure is the topmost structural part of a reactor module. For a pool-type fast reactor it is an especially important structure because it provides the interface between the primary coolant system and the main access area above the closure. The reactor closure comprises a stationary deck, a rotatable plug, the boundary elements of primary system and containment penetrations for equipment and auxiliary systems. This paper evaluates two different reactor closure design concepts, referred to as ''warm'' deck and ''hot'' deck, for a pool-type fast reactor with respect to their design features, technical merits, and economic benefits. The evaluation also includes functional, structural, and thermal analyses of the two deck design concepts. Issues related to their fabrication and shipping to the plant site are also addressed. The warm deck is a thick solid steel plate with under-the-deck insulation consisting of many layers of steel plates. The hot deck is a box-type structure consisting of a bottom plate reinforced with vertical ribs and cylinders. For insulation and radiation shielding, the region of the hot deck above the bottom plate is filled with steel balls. Conventional insulation is added on the top to further reduce heat loss into area above the deck. The design choice of the closure deck is strongly dependent on design features of the reactor; especially on the reactor module support. While the warm deck is preferable with the top support, the hot deck is better suited for the bottom support design of the module.

Chung, H.; Seidensticker, R.W.; Kann, W.J.; Bump, T.R.; Schatmeier, C.

1986-01-01T23:59:59.000Z

31

W. Alexander Williams, Ph.D Designation and Certification Manager  

Office of Legacy Management (LM)

March 20, 1995 March 20, 1995 W. Alexander Williams, Ph.D Designation and Certification Manager EM-421 Quince Orchard U.S. Department of Energy Washington, DC 20585-0002 L -=kj SUBJECT: FINAL REPORT-RADIOLOGICAL SURVEY OF THE F AMERICAN MANUFACTURING COMPANY OF TEXA WORTH, TEXAS I ,d Sl I lRMER , FORT Dear Dr. Williams: The Environmental,Survey and Site Assessment Program (ESSAP) of the C for Science and Education (ORISE) conducted radiological survey activii American Manufacturing Company of Texas Site in Fort Worth, Texas October 3 through 5, 1994. The survey activities consisted of document revi surface activity measurements, smears for removable activity, exposure rate m soil and miscellaneous sampling. k Ridge Institute :s at the former wing the period

32

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network (OSTI)

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

33

Reactor protection system design alternatives for sodium fast reactors  

E-Print Network (OSTI)

Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

DeWitte, Jacob D. (Jacob Dominic)

2011-01-01T23:59:59.000Z

34

Fusion reactor design studies. [ARIES Tokamak  

SciTech Connect

This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)

Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

1990-10-12T23:59:59.000Z

35

Advanced Burner Test Reactor - Preconceptual Design Report  

NLE Websites -- All DOE Office Websites (Extended Search)

Burner Test Reactor Preconceptual Design Report ANL-ABR-1 (ANL-AFCI-173) Nuclear Engineering Division Disclaimer This report was prepared as an account of work sponsored by an...

36

Conceptual design study of spheromak reactors  

SciTech Connect

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain characteristic relations among various parameters of the spheromak configuration for an aspect ratio A greater than or equal to 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) DT two-component, (2) DT ignited, and, (3) catalyzed DD ignited reactors. With a total wall loading of approx. 4 MWm/sup -2/, it is found that edge magnetic fields of only approx. 4T (DT) and approx. 9T (cat. DD) are required for ignited reactors of one-meter plasma (minor) radius with output powers in the gigawatt range. Assessment of various methods of generating reactor-grade spheromak plasmas is discussed briefly.

Katsurai, M.; Yamada, M.

1980-07-01T23:59:59.000Z

37

Mirror Advanced Reactor Study interim design report  

DOE Green Energy (OSTI)

The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

Not Available

1983-04-01T23:59:59.000Z

38

Generic small modular reactor plant design.  

SciTech Connect

This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

2012-12-01T23:59:59.000Z

39

CONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV  

E-Print Network (OSTI)

step and present large scale tokamak design groups should be encouraged. 3. The next reactor design (8th April 1984). These devices are the so called 'Large Tokamaks', because their parameters (Table 1 AND SYMPOSIA TABLE 1-1. CURRENT DATA BASE FOR LARGE TOKAMAKS Parameter/data TFTR JET JT-6O T-15 Major radius (m

Abdou, Mohamed

40

Design options for a bunsen reactor.  

SciTech Connect

This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

Moore, Robert Charles

2013-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Basic and Applied Science Research Reactors - Reactors designed...  

NLE Websites -- All DOE Office Websites (Extended Search)

BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...

42

Computer simulations help design new nuclear reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Computer simulations help design new nuclear reactors Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Reprinted from "Argonne Now" - Spring 2008 Physicist Won-Sik Yang and computer scientist Andrew Siegel hold a fuel rod assembly in front of a model of the Experimental Breeder Reactor-II

43

The Argonaut Reactor - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

44

Thermal and flow design of helium-cooled reactors  

Science Conference Proceedings (OSTI)

This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors.

Melese, G.; Katz, R.

1984-01-01T23:59:59.000Z

45

Design studies of mirror machine reactors  

SciTech Connect

An overview is presented of a mirror fusion reactor design study. The general methodology used in the study is discussed, the reactor is described, and some design alternatives to the present approach are enumerated. The system chosen for this design study is a mirror machine with direct conversion using D- T fuel. The nominal power output is 200 MW. The coil geometry is the Yin Yang, minimum B with a vacuum mirror ratio of 3. The coil is of particular utility because of its simple conductor shapes and because the two separate conductors, by proper B-field biasing, allow the charged particles to escape preferentially through one mirror only and through a relatively small window'' of that mirror. This is necessary for direct converter economy. (auth)

Werner, R.W.; Carlson, G.A.; Hovingh, J.; Lee, J.D.; Peterson, M.A.

1973-12-01T23:59:59.000Z

46

Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 1, Chapters 1--14  

Science Conference Proceedings (OSTI)

This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE`s System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report.

Not Available

1994-08-01T23:59:59.000Z

47

Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices  

SciTech Connect

This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE`s System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E.

Not Available

1994-08-01T23:59:59.000Z

48

Innovative design of uranium startup fast reactors  

E-Print Network (OSTI)

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

49

Fusion-breeder-reactor design studies  

SciTech Connect

Studies of the technical and economic feasibility of producing fissile fuel in tandem mirrors and in tokamaks for use in fission reactors are presented. Fission-suppressed fusion breeders promise unusually good safety features and can provide make-up fuel for 11 to 18 LWRs of equal nuclear power depending on the fuel cycle. The increased revenues from sales of both electricity and fissile material might allow the commercial application of fusion technology significantly earlier than would be possible with electricity production from fusion alone. Fast-fission designs might allow a fusion reactor with a smaller fusion power and lower Q value to be economical and thus make this application of fusion even earlier. A demonstration reactor with a fusion power of 400 MW could produce 600 kg of fissile material per year at a capacity factor of 50%. The critical issues, for which small scale experiments are either being carried out or planned, are: (1) material compatibility, (2) beryllium feasibility, (3) MHD effects, and (4) pyrochemical reprocessing.

Moir, R.W.; Lee, J.D.; Coops, M.S.

1983-04-05T23:59:59.000Z

50

Neutron transport analysis for nuclear reactor design  

DOE Patents (OSTI)

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

Vujic, J.L.

1993-11-30T23:59:59.000Z

51

Neutron transport analysis for nuclear reactor design  

DOE Patents (OSTI)

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

Vujic, Jasmina L. (Lisle, IL)

1993-01-01T23:59:59.000Z

52

Advanced High Temperature Reactor Neutronic Core Design  

Science Conference Proceedings (OSTI)

The AHTR is a 3400 MW(t) FHR class reactor design concept intended to serve as a central generating station type power plant. While significant technology development and demonstration remains, the basic design concept appears sound and tolerant of much of the remaining performance uncertainty. No fundamental impediments have been identified that would prevent widespread deployment of the concept. This paper focuses on the preliminary neutronic design studies performed at ORNL during the fiscal year 2011. After a brief presentation of the AHTR design concept, the paper summarizes several neutronic studies performed at ORNL during 2011. An optimization study for the AHTR core is first presented. The temperature and void coefficients of reactivity are then analyzed for a few configurations of interest. A discussion of the limiting factors due to the fast neutron fluence follows. The neutronic studies conclude with a discussion of the control and shutdown options. The studies presented confirm that sound neutronic alternatives exist for the design of the AHTR to maintain full passive safety features and reasonable operation conditions.

Ilas, Dan [ORNL; Holcomb, David Eugene [ORNL; Varma, Venugopal Koikal [ORNL

2012-01-01T23:59:59.000Z

53

Heavy Water and Graphite Reactors - Reactors designed/built by...  

NLE Websites -- All DOE Office Websites (Extended Search)

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

54

Manhattan Project: Production Reactor (Pile) Design, Met Lab, 1942  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN (Met Lab, 1942) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 By 1942, scientists had established that some of the uranium exposed to radioactivity in a reactor (pile) would eventually decay into plutonium, which could then be separated by chemical means from the uranium. Important theoretical research on this was ongoing, but the work was scattered at various universities from coast to coast. In early 1942, Arthur Compton arranged for all pile research to be moved to the Met Lab at the University of Chicago.

55

Evaluation of the economic simplified boiling water reactor human reliability analysis using the SHARP framework  

E-Print Network (OSTI)

General Electric plans to complete a design certification document for the Economic Simplified Boiling Water Reactor to have the new reactor design certified by the United States Nuclear Regulatory Commission. As part of ...

Dawson, Phillip Eng

2007-01-01T23:59:59.000Z

56

HIGH FLUX ISOTOPE REACTOR PRELIMINARY DESIGN STUDY  

SciTech Connect

A comparison of possible types of research reactors for the production of transplutonium elements and other isotopes indicates that a flux-trap reactor consisting of a beryllium-reflecteds light-water-cooled annular fuel region surrounding a light-water island provides the required thermal neutron fluxes at minimum cost. The preliminary desigu of such a reactor was carried out on the basis of a parametric study of the effect of dimensions of the island and fuel regions heat removal rates, and fuel loading on the achievable thermal neutmn fluxes in the island and reflector. The results indicate that a 12- to 14-cm- diam. island provides the maximum flux for a given power density. This is in good agreement with the US8R critical experiments. Heat removal calculations indicate that average power densities up to 3.9 Mw/liter are achievable with H/ sub 2/O-cooled, platetype fuel elements if the system is pressurized to 650 psi to prevent surface boiling. On this basis, 100 Mw of heat can be removed from a 14-cm-ID x 36-cm-OD x 30.5-cm-long fuel regions resulting in a thermal neutron flux of 3 x 10/sup 15/ in the island after insertion of 100 g of Cm/sup 244/ or equivalent. The resulting production of Cf/sup 252/ amounts to 65 mg for a 1 1/2- year irradiation. Operation of the reactor at the more conservative level of 67 Mw, providing an irradiation flux of 2 x 10/sup 15/ in the islands will result in the production of 35 mg of Cf/sup 252/ per 18 months from 100 g of Cm/sup 244/. A development program is proposed to answer the question of the feasibility of the higher power operation. In addition to the central irradiation facility for heavyelement productions the HFIR contains ten hydraulic rabbit tubes passing through the beryllium reflector for isotope production and four beam holes for basic research, Preliminary estimates indicate that the cost of the facility, designed for an operating power level of 100 Mw, will be approximately 2 million. (auth)

Lane, J.A.; Cheverton, R.D.; Claiborne, G.C.; Cole, T.E.; Gambill, W.R.; Gill, J.P.; Hilvety, N.; McWherther, J.R.; Vroom, D.W.

1959-03-20T23:59:59.000Z

57

Space Reactor Radiation Shield Design Summary, for Information  

SciTech Connect

The purpose of this letter is to provide a summary of the Prometheus space reactor radiation shield design status at the time of program restructuring.

EC Pheil

2006-02-17T23:59:59.000Z

58

Updated Uranium Fuel Cycle Environmental Impacts for Advanced Reactor Designs  

Science Conference Proceedings (OSTI)

The purpose of this project was to update the environmental impacts from the uranium fuel cycle for select advanced (GEN III+) reactor designs.

Nitschke, R.

2004-10-03T23:59:59.000Z

59

Reactor Design for CO2 Capture Using Sorbents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Design for CO 2 Capture Using Sorbents Background Carbon Sequestration is rapidly becoming accepted as a viable option to reduce the amount of carbon dioxide (CO 2 )...

60

Design considerations in inertially-confined fusion reactors  

SciTech Connect

This paper discusses the effects of short time pulses of energetic particles and waves typical of inertially-confined thermonuclear reactions on the first wall, blanket and shield of conceptual reactors. Several reactor designs are presented which attempt to cope with the various problems from the microexplosion debris. Fusion-fission hybrid reactors are also discussed. Emphasis is placed on the first-wall problems of laser-initiated, inertially confined fusion reactors using the deuterium-tritium fuel cycle.

Hovingh, J.

1976-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Revised design for the Tokamak experimental power reactor  

DOE Green Energy (OSTI)

A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost.

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1977-03-01T23:59:59.000Z

62

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs The Small Modular Reactor (SMR) Licensing Technical Support (LTS) program, sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), through this Funding Opportunity Announcement (FOA) seeks to facilitate the development of innovative SMR designs that have the potential to address the nation's economic, environmental and energy security goals. Specifically, the Department is soliciting applications for SMR designs that offer unique and innovative solutions for achieving the objectives of enhanced safety, operations, and performance relative to currently certified designs. This FOA focuses on design development and

63

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs The Small Modular Reactor (SMR) Licensing Technical Support (LTS) program, sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), through this Funding Opportunity Announcement (FOA) seeks to facilitate the development of innovative SMR designs that have the potential to address the nation's economic, environmental and energy security goals. Specifically, the Department is soliciting applications for SMR designs that offer unique and innovative solutions for achieving the objectives of enhanced safety, operations, and performance relative to currently certified designs. This FOA focuses on design development and

64

Design guide for Category III reactors: pool type reactors. [US DOE  

SciTech Connect

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

1978-11-01T23:59:59.000Z

65

Summary of the MARS tandem-mirror reactor design  

SciTech Connect

A recently completed two-year study of a commercial tandem-mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped-particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24 T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted. General characteristics of the MARS tandem mirror and STARFIRE tokamak reactor design are compared. A design of an upgrade of MFTF-B incorporating many of the MARS features is discussed.

Logan, B.G.

1983-09-01T23:59:59.000Z

66

Design Considerations for Economically Competitive Sodium Cooled Fast Reactors  

SciTech Connect

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

Hongbin Zhang; Haihua Zhao

2009-05-01T23:59:59.000Z

67

Transpiring wall supercritical water oxidation test reactor design report  

Science Conference Proceedings (OSTI)

Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G. [Sandia National Labs., Livermore, CA (United States). Engineering for Transportation and Environment Dept.; Rousar, D.C. [GenCorp Aerojet, Sacramento, CA (United States)

1996-02-01T23:59:59.000Z

68

ALPR PRELIMINARY DESIGN STUDY (ARGONNE LOW POWER REACTOR). PHASE I  

SciTech Connect

A preliminary design study, Phase I of the ALPR . project, has been made in accordance with the Army Reactors Branch specifications for a nuclear ''package'' power plant with a 200 to 260 kw electric and 400 kw heating capacity..The plant is to be installed at the Idaho Reactor Testing Station as a prototype for remote arctic installations. The ''conventiornl'' power plant as well as the exterior reactor components are described, and cost estimates are given. ''Nuclear'' components of the reactor are described. (auth)

Treshow, M.; Hamer, E.; Pearlman, H.; Rossin, D.; Shaftman, D.

1956-04-20T23:59:59.000Z

69

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

70

Design guide for category VI reactors: air-cooled graphite reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned air-cooled graphite reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC).

Brynda, W.J.; Karol, R.; Powell, R.W.

1979-02-01T23:59:59.000Z

71

System 80+{trademark} standard design: CESSAR design certification. Volume 15: Amendment I  

Science Conference Proceedings (OSTI)

This report has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{trademark} Standard Design.

Not Available

1990-12-21T23:59:59.000Z

72

NGNP Nuclear-Industrial Facility and Design Certification Boundaries White Paper  

SciTech Connect

The Next Generation Nuclear Plant (NGNP) Project was initiated at Idaho National Laboratory by the U.S. Department of Energy pursuant to the 2005 Energy Policy Act and based on research and development activities supported by the Generation IV Nuclear Energy Systems Initiative. The principal objective of the NGNP Project is to support commercialization of the high temperature gas-cooled reactor (HTGR) technology. The HTGR is helium cooled and graphite moderated and can operate at reactor outlet temperatures much higher than those of conventional light water reactor (LWR) technologies. Accordingly, it can be applied in many industrial applications as a substitute for burning fossil fuels, such as natural gas, in addition to producing electricity, which is the principal application of current LWRs. These varied industrial applications may involve a standard HTGR modular design using different Energy Conversion Systems. Additionally, some of these process heat applications will require process heat delivery systems to lie partially outside the HTGR operator’s facility.

Thomas E. Hicks

2011-07-01T23:59:59.000Z

73

Performance and safety design of the advanced liquid metal reactor  

SciTech Connect

The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel.

Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L. (General Electric Co., San Jose, CA (United States). Advanced Nuclear Technology)

1991-01-01T23:59:59.000Z

74

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors  

Science Conference Proceedings (OSTI)

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01T23:59:59.000Z

75

Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor  

E-Print Network (OSTI)

A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated ...

Hejzlar, P.

76

Liquid Metal Fast Breeder Reactor plant maintenance and equipment design  

Science Conference Proceedings (OSTI)

This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment.

Swannack, D.L.

1982-06-07T23:59:59.000Z

77

A DESCRIPTION OF INTEGRAL PHYSICS DATA FOR FAST REACTOR DESIGN  

SciTech Connect

Integral physics data for fast reactor design are discussed. The measurements needed include those of critical mass, shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi- alpha , and similar quantities. Topics covered include Pu- and U/sup 233/-fueled systems, highly enriched U/sup 235/ systems in optimum geometry, uranium cores of various enrichments and dilutions, extreme geometry critical experiments, specific reactor systems, core mockup inhomogeneities, spectral studies and detector ratios, uranium equilibrium spectrum data, materialreplacement measurements, fast reactor dynamics, and suggested future experiments and experimental programs. (M.C.G.)

Loewenstein, W.B.; Meneghetti, D.

1961-07-01T23:59:59.000Z

78

AN 80 MEGAWATT AQUEOUS HOMOGENEOUS BURNER REACTOR. Reactor Design and Feasibility Problem  

SciTech Connect

An 80 Mw aqueous homogeneous burner reactor suitable for producing 20 Mw of electricity at a remote location is described. The reactor fuel consists of a light water uranyl sulfate solution which acts as its own moderator and coolant. The uranium is highly enriched (93% U/sup 235/). The primary considerstions for the design were simplicity and reliability of the components, automatic demand control and safe for any load change, full xenon override not required, possibility of construction within the immediate future, and economic operation not the cortrolling factor. Reasonably complete studies are presented for the reactor physics, safety, stability, chemistry, hent transfer, and operation of the system. (auth)

Chapman, R.H.; Collins, H.L.; Dollard, W.J.; Fieno, D.; Hernandez- Fragoso, J.; Miller, J.W.; von Hollen, H.; Wheeler, C.V.

1957-08-01T23:59:59.000Z

79

THE ADVANCED TEST REACTOR-ATR FINAL CONCEPTUAL DESIGN  

SciTech Connect

The results of a study are presented which provided additional experimental-loop irradiation space for the AECDRD testing program. It was a premise that the experiments allocated to this reactor were those which could not be accommodated in the MTR, ETR, or in existing commercial test reactors. To accomplish the design objectives called for a reactor producing perturbed neutron fluxes exceeding 1O/sup 15/ thermal n/cm/sup 2/-sec and 1.5 x 1O/sup 15/ epithermal n/cm/sup 2/-sec. To accommodate the experimental samples, the reactor fuel core is four feet long in the direction of experimental loops. This is twice the length of the MTR core and a third longer than the ETR core. The vertical arrangement of reactor and experiments permits the use of loops penetrating the top cap of the reactor vessel running straight and vertically through the reactor core. The design offers a high degree of accessibility of the exterior portions of the experiments and offers very convenient handling and discharge of experiments. Since the loops are to be integrated into the reactor design and the in-pile portions installed before reactor start-up, it is felt that many of the problems encountered in MTR and ETR experience will cease to exist. Installation of the loops prior to startup will have an added advantage in that the flux variations experienced in experiments in ETR every time a new loop is installed will be absent. The Advanced Test Reactor has a core configuration that provides essentially nine flux-trap regions in a geometry that is almost optimum for cylindrical experiments. The geometry is similar to that of a fourleaf clover with one flux trap in each leaf, one at the intersection of the leaves, and one between each pair of leaves. The nominal power level is 250 Mw. The study was carried out in enough detail to permit the establishment of the design parameters and to develop the power requirement which, conservatively rated, will definitely reach the flux specifications. A critical mockup of an arrangement similar to ATR was loaded into the Engineering Test Reactor Critical Facility. (auth)

deBoisblanc, D.R. et al

1960-11-01T23:59:59.000Z

80

Basis for NGNP Reactor Design Down-Selection  

Science Conference Proceedings (OSTI)

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

L.E. Demick

2010-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Basis for NGNP Reactor Design Down-Selection  

SciTech Connect

The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

L.E. Demick

2011-11-01T23:59:59.000Z

82

Design, fabrication, and certification of advanced modular PV power systems. Final technical progress report  

DOE Green Energy (OSTI)

Solar Electric Specialties Company (SES) has completed a two and a half year effort under the auspices of the US Department of Energy (DOE) PVMaT (Photovoltaic Manufacturing Technology) project. Under Phase 4A1 of the project for Product Driven System and Component Technology, the SES contract ``Design, Fabrication and Certification of Advanced Modular PV Power Systems`` had the goal to reduce installed system life cycle costs through development of certified (Underwriters Laboratories or other listing) and standardized prototype products for two of the product lines, MAPPS{trademark} (Modular Autonomous PV Power Supply) and Photogensets{trademark}. MAPPS are small DC systems consisting of Photovoltaic modules, batteries and a charge controller and producing up to about a thousand watt-hours per day. Photogensets are stand-alone AC systems incorporating a generator as backup for the PV in addition to a DC-AC inverter and battery charger. The program tasks for the two-year contract consisted of designing and fabricating prototypes of both a MAPPS and a Photogenset to meet agency listing requirements using modular concepts that would support development of families of products, submitting the prototypes for listing, and performing functionality testing at Sandia and NREL. Both prototypes were candidates for UL (Underwriters Laboratories) listing. The MAPPS was also a candidate for FM (Factory Mutual) approval for hazardous (incendiary gases) locations.

Lambarski, T.; Minyard, G. [Solar Electric Specialties Co., Willits, CA (United States)

1998-10-01T23:59:59.000Z

83

MHD equilibrium properties of tokamak fusion reactor designs  

SciTech Connect

The equilibrium properties of several Tokamak Reactor Designs are analyzed and compared for varying pressure and current profiles using the Princeton Equilibrium Code. It is found that the UWMAK configuration has a broader range of equilibria than the Princeton Reference Design configuration, but that the safety factor on axis is less than unity for peaked current distributions. The Argonne Experimental Power Reactor has a satisfactory range of equilibria, but a means of limiting or diverting the plasma has not yet been proposed, and this may substantially change the results obtained. (auth)

Todd, A. M.M.; Gralnick, S. L.; Dalhed, H. E.

1976-01-01T23:59:59.000Z

84

DESIGN OF A MICROCHANNEL BASED SOLAR RECEIVER/REACTOR FOR  

E-Print Network (OSTI)

DESIGN OF A MICROCHANNEL BASED SOLAR RECEIVER/REACTOR FOR METHANE-STEAM REFORMING Drost, K. J- current experimental data [1, 2]. Methane-steam reforming is modeled by three reduced-order reactions and modeling are used to investigate the strong endothermic reactions of methane-steam reforming inside

Apte, Sourabh V.

85

Constructal method to optimize solar thermochemical reactor design  

SciTech Connect

The objective of this study is the geometrical optimization of a thermochemical reactor, which works simultaneously as solar collector and reactor. The heat (concentrated solar radiation) is supplied on a small peripheral surface and has to be dispersed in the entire reactive volume in order to activate the reaction all over the material. A similarity between this study and the point to volume problem analyzed by the constructal approach (Bejan, 2000) is evident. This approach was successfully applied to several domains, for example for the coupled mass and conductive heat transfer (Azoumah et al., 2004). Focusing on solar reactors, this work aims to apply constructal analysis to coupled conductive and radiative heat transfer. As a first step, the chemical reaction is represented by a uniform heat sink inside the material. The objective is to optimize the reactor geometry in order to maximize its efficiency. By using some hypothesis, a simplified solution is found. A parametric study provides the influence of different technical and operating parameters on the maximal efficiency and on the optimal shape. Different reactor designs (filled cylinder, cavity and honeycomb reactors) are compared, in order to determine the most efficient structure according to the operating conditions. Finally, these results are compared with a CFD model in order to validate the assumptions. (author)

Tescari, S.; Mazet, N. [PROMES-CNRS, Rambla de la Thermodynamique, Tecnosud, 66100 Perpignan (France); Neveu, P. [PROMES-CNRS, Rambla de la Thermodynamique, Tecnosud, 66100 Perpignan (France); Universite de Perpignan Via Domitia, 52 Avenue Paul Alduy, 66860 Perpignan (France)

2010-09-15T23:59:59.000Z

86

Reactor core design and modeling of the MIT research reactor for conversion to LEU  

SciTech Connect

Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

Newton, Thomas H. Jr. [Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Olson, Arne P.; Stillman, John A. [RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

2008-07-15T23:59:59.000Z

87

Core design and reactor physics of a breed and burn gas-cooled fast reactor  

E-Print Network (OSTI)

In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle ...

Yarsky, Peter

2005-01-01T23:59:59.000Z

88

Mechanical design of a light water breeder reactor  

DOE Patents (OSTI)

In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

Fauth, Jr., William L. (Germantown, MD); Jones, Daniel S. (Pittsburgh, PA); Kolsun, George J. (Pittsburgh, PA); Erbes, John G. (San Jose, CA); Brennan, John J. (Bethel Park, PA); Weissburg, James A. (Pittsburgh, PA); Sharbaugh, John E. (Acme, PA)

1976-01-01T23:59:59.000Z

89

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor  

Science Conference Proceedings (OSTI)

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)

Poston, David I. [Nuclear Systems Design Group, Decision Applications Division, Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States)

2002-07-01T23:59:59.000Z

90

Thermal hydraulic design of a salt-cooled highly efficient environmentally friendly reactor  

E-Print Network (OSTI)

A 1 OOOMWth liquid-salt cooled thermal spectrum reactor was designed with a long fuel cycle, and high core exit temperature. These features are desirable in a reactor designed to provide process heat applications such as ...

Whitman, Joshua (Joshua J.)

2009-01-01T23:59:59.000Z

91

Design of passive decay heat removal system for the lead cooled flexible conversion ratio fast reactor  

E-Print Network (OSTI)

The lead-cooled flexible conversion ratio fast reactor shows many benefits over other fast-reactor designs; however, the higher power rating and denser primary coolant present difficulties for the design of a passive decay ...

Whitman, Joshua (Joshua J.)

2007-01-01T23:59:59.000Z

92

FAST FUEL TEST REACTOR-FFTR CONCEPTUAL DESIGN STUDY  

SciTech Connect

The Fast Fuel Test Reactor (FFTR) is a nuclear facility for the purpose of irradiating samples of fuels and structural components for use in fast reactors. The core consisis of a plate type element in a square configuration. Beryllium metal between the fuel elements is used to obtain a neutron energy spectrum in the hard intermediate region. Cooling of the core and test specimens is accomplished by means of liquid sodium. The design concept was carried through in sufficient degree in the following areas of preliminary concern: number and size of irradiation facilities, sample power requirements, plant layout to evaluate site requirements, plant and nuclear design parameters to evaluate essential equipment requirements. plant-capital-cost estimate, annual- operating-cost estimate, and estimate of construction time schedule. (W.D.M.)

Brubaker, R.; Hummel, H.H.; McArthy, A.; Smaardyk, A.; Kittel, J.H.

1960-08-01T23:59:59.000Z

93

Novel Reactor Design for Solid Fuel Chemical Looping Combustion  

NLE Websites -- All DOE Office Websites (Extended Search)

Novel Reactor Design for Solid Fuel Novel Reactor Design for Solid Fuel Chemical Looping Combustion Opportunity Research is active on the patent pending technology, titled "Apparatus and Method for Solid Fuel Chemical Looping Combustion." This technology is available for licensing and/or further collaborative research from the U.S. Department of Energy's National Energy Technology Laboratory. Overview The removal of CO2 from power plants is challenging because existing methods to separate CO2 from the gas mixture requires a significant fraction of the power plant output. Chemical-looping combustion (CLC) is a novel technology that utilizes a metal oxide oxygen carrier to transport oxygen to the fuel thereby avoiding direct contact between fuel and air. The use of CLC has the advantages of reducing the energy penalty while

94

PRA insights applicable to the design of the Broad Applications Test Reactor  

SciTech Connect

Design insights applicable to the design of a new Broad Applications Test Reactor (BATR), being studied at Idaho National Engineering Laboratory, are summarized. Sources of design insights include past probabilistic risk assessments and related studies for department of Energy-owned Class A reactors and for commercial reactors. The report includes a preliminary risk allocation scheme for the BATR.

Khericha, S.T.; Reilly, H.J.

1993-01-01T23:59:59.000Z

95

GAS COOLED PEBBLE BED REACTOR FOR A LARGE CENTRAL STATION. Reactor Design and Feasibility Study  

SciTech Connect

An optimum econonic design for a high temperature, helium cooled, central station reactor power plant of about 400 Mw of electric power was determined. The core consists of a randomly packed bed of unclad graphite spheres, approximately one in. in diameter, impregnated with U/sup 233/ and thorium such that a conversion ratio of near unity is achieved. The high temperature helium permits steam conditions, at the turbine throttle, of 1000 deg F and 1450 psia. (auth)

Schock, A.; Bruley, D.F.; Culver, H.N.; Ianni, P.W.; Kaufman, W.F.; Schmidt, R.A.; Supp, R.E.

1957-08-01T23:59:59.000Z

96

Neutronic and thermal design considerations for heat-pipe reactors  

SciTech Connect

SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 1500/sup 0/K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined.

Ranken, W.A.; Koenig, D.R.

1983-01-01T23:59:59.000Z

97

CONCEPTUAL DESIGN OF THE PEBBLE BED REACTOR EXPERIMENT  

SciTech Connect

The Pebble-Bed Reactor Experiment (PBRE) was designed to advance the pebble-bed concept by providing a test of characteristic features and make contriliutions to the general development of all-ceramic gas-cooled reactors. The following objectives were established for the reactor experiment: to investigate key features of the pebble-bed concept, including on-stream fuel handling, movement of fuel through bed, and performance of core; to obtain operation and maintenance experience with a system contaminated with fission- product activity; and to investigate the behavior of graphite fuel elements. A fourth objective, study of the behavior of core materials at conditions occurring with exit gas temperatures in the range 2000 to 2500 deg F, was tentatively included. The preliminary design oE a 5-Mw(t) reactor for achieving these objectives was prepared. The core of the PBRE is a 2 1/2-ft-diam, 4-ft-tall cylinder containing approximately 12,000 spherical graphite fuel elements 1 1/2 in. in diameter. Fuel spheres are added to and removed from the core by gravity flow, and these operations are performed while the reactor is at power by using pairs of valves for passage of elements into and out of the high-pressure system. Exposed fuel can be recycled to the top of the core. Helium coolant at 500 psia enters the bottom of the core at 550 deg F and emerges from the top at 1250 deg F. Concentric ducting connects the reactor to a single heat exchanger, which is located sufficiently high above the core that natural circulation will suffice to remove afterheat in the event the blower ceases to function. The coolant flow path is such that the entire pressure envelope is swept with helium at the temperature at which it emerges from the heat exchanger. Provision for semi- remote maintenance of contaminated components is emphasized in the layout, and most of the equipment in the primary and auxiliary systems is accessible from above by the removal of modular shielding units. Thc design permits replacement of the entire core graphite structure, The reactor can be adapted for testing core materials at high temperature by attemperation of the hot helium emerging from the core wwiih cool gas in a plenum in the upper graphite structure. Location of the PBRE at the site of the HRE-2 facility is proposed to take advantage of available buildings and services, but the reactor and auxiliary equipment will be contained in a completely new vessel located adjacent to the existing building. The design and direct construction cost of the reactor plant is estimated to be 958,000, allowance for contingencies, overhead, and escalation brings the total to ,260,000. High-temperature operation can be achieved when desired for an additional expenditure of less than 0,000. (auth)

1962-05-17T23:59:59.000Z

98

Design of an LEU core for the MIT reactor  

Science Conference Proceedings (OSTI)

A design of the MIT Reactor core using monolithic U-7Mo LEU fuel has been developed with the goal of maintaining thermal and fast neutron fluxes as well as increasing the flexibility for meeting the needs of in-core experiments. An optimum core was sought by varying the core materials, and fuel plate numbers and thicknesses, but maintaining the outside dimensions of a fuel element. A full-core model of the MITR by the Monte-Carlo transport code MCNP was used to calculate the neutron fluxes, reactivity and neutron spectrum available for experiments. The optimum reactor design consisted of the use of half-sized fuel elements made up of nine U-7Mo LEU fuel plates of 0.55 mm thickness with 0.25 mm finned aluminum cladding. This design also utilized solid beryllium fuel elements (dummies) with boron fixed absorbers or solid lead dummies, depending on the in-core experiment flux and spectrum needs. Because the new core design contains twice the amount of 235 U as does the existing HEU core, and produces much more Pu, its fuel cycle length is twice as long at the same power level. Preliminary thermal-hydraulic and neutronic safety evaluations indicate superior performance to the current HEU fuel. (authors)

Newton, T. [Massachusetts Inst. of Technology, Nuclear Reactor Laboratory, 138 Albany St., Cambridge, MA 02139 (United States); Kazimi, M.; Pilat, E. [Nuclear Science and Engineering Dept., 77 Massachusetts Ave., Cambridge, MA 02139 (United States)

2006-07-01T23:59:59.000Z

99

Achievements: Nuclear Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne National Laboratory Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

100

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents (OSTI)

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Huebotter, P.R.; McLennan, G.A.

1984-08-30T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents (OSTI)

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Huebotter, Paul R. (Western Springs, IL); McLennan, George A. (Downers Grove, IL)

1985-01-01T23:59:59.000Z

102

Design, optimization and evaluation of a free-fall biomass fast pyrolysis reactor and its products.  

E-Print Network (OSTI)

??The focus of this work is a radiatively heated, free-fall, fast pyrolysis reactor. The reactor was designed and constructed for the production of bio-oil from… (more)

Ellens, Cody James

2009-01-01T23:59:59.000Z

103

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

104

Early Exploration - Reactors designed/built by Argonne National Laboratory  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Exploration Early Exploration About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

105

Seismic responses of a pool-type fast reactor with different core support designs  

Science Conference Proceedings (OSTI)

In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs.

Wu, Ting-shu; Seidensticker, R.W. (Argonne National Lab., IL (USA))

1989-01-01T23:59:59.000Z

106

Preconceptual design and assessment of a Tokamak Hybrid Reactor  

SciTech Connect

The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

1980-09-01T23:59:59.000Z

107

DESIGN OF THE ARGONNE LOW POWER REACTOR (ALPR)  

SciTech Connect

A description is given of the design of a prototype "packaged" nuclear power plant. The purpose of the plant is to alleviate fuel oil logistics and storage problems posed by remote auxiliary DEW Line radar statibns north of the Arctic Circle. The ALPR (redesignated SL-1) is a 3 Mwt, heterogeneous, highly enriched uranium- fueled, naturalcirculation boiling water reactor, ccoled and moderated with light water. Steam at 300 psig, dry and saturated (421 deg F) is passed directly from the reactor to a conventional turbine-generator to produce electric power (300 kw nominal) and space-heating (400 kw) requirements consistent with rigid mechanical and structural specifications prescribed by the military, and dictated by the extreme geophysics prevailing at the ultimate site. The over all design criteria emphasize: simplicity and reliability of operation and maintenance, with minimum supervision; minimum on-site construction; maximum use of standard components; limited water supply; utilization of local gravel for biological shielding; transportability by air lift; and nominal 3-year fuel operating lifetime per core loading. The "packaged" concept is incorporated for the initial erection. The plant is not designed for relocation. The design criteria for the prototype necessitate special features. The fuel plates are clad with an alurninurn-nickel alloy (X8001). Burnable-poison (BIO) strips are mechancally attached to the fuel assemblies to compensate the excess reactivity required for a nominal 3-year core operating lifetime. The control rods are actuated by rackand-pinion drive extensions which incorporate rotary seals. Fuel exchange is accomplished without the removal of the pressure vessel head. The electrical power generated is used to operate plant auxiliaries; the "net electric power" is dissipated by resistors. The hot water for space heating is heated in a heat exchanger by 20-psig steam, use being made of the latent heat of vaporization, and all the heat is dissipated by a finned-tube, air-cooled heat exchanger. (auth)

Grant, N.R.; Hamer, E.E.; Hooker, H.H.; Jorgensen, G.L.; Kann, W.J.; Lipinski, W.C.; Milak, G.C.; Rossin, A.D.; Shaftman, D.H.; Smaardyk, A.; Treshow, M.

1961-05-01T23:59:59.000Z

108

22.39 Integration of Reactor Design, Operations, and Safety, Fall 2005  

E-Print Network (OSTI)

This course integrates studies of reactor physics and engineering sciences into nuclear power plant design. Topics include materials issues in plant design and operations, aspects of thermal design, fuel depletion and ...

Todreas, Neil E.

109

DOI Designates B Reactor at DOE's Hanford Site as a National Historic  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOI Designates B Reactor at DOE's Hanford Site as a National DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark August 25, 2008 - 3:20pm Addthis DOE to offer regular public tours in 2009 WASHINGTON, DC - U.S. Department of the Interior (DOI) Deputy Secretary Lynn Scarlett and U.S. Department of Energy (DOE) Acting Deputy Secretary Jeffrey F. Kupfer today announced the designation of DOE's B Reactor as a National Historic Landmark and unveiled DOE's plan for a new public access program to enable American citizens to visit B Reactor during the 2009 tourist season. The B Reactor at DOE's Hanford Site in southeast Washington State was the world's first industrial-scale nuclear reactor and produced plutonium for the atomic weapon that was dropped on Nagasaki,

110

DOI Designates B Reactor at DOE's Hanford Site as a National Historic  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOI Designates B Reactor at DOE's Hanford Site as a National DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark August 25, 2008 - 3:20pm Addthis DOE to offer regular public tours in 2009 WASHINGTON, DC - U.S. Department of the Interior (DOI) Deputy Secretary Lynn Scarlett and U.S. Department of Energy (DOE) Acting Deputy Secretary Jeffrey F. Kupfer today announced the designation of DOE's B Reactor as a National Historic Landmark and unveiled DOE's plan for a new public access program to enable American citizens to visit B Reactor during the 2009 tourist season. The B Reactor at DOE's Hanford Site in southeast Washington State was the world's first industrial-scale nuclear reactor and produced plutonium for the atomic weapon that was dropped on Nagasaki,

111

The Integral Fast Reactor (IFR) - Reactors designed/built by Argonne  

NLE Websites -- All DOE Office Websites (Extended Search)

Integral Fast Reactor Integral Fast Reactor About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

112

lullIIlllIllLLLII DESIGN WINDOWS FOR A He COOLED FUSION REACTOR*  

E-Print Network (OSTI)

_ ii|l iImMmmm lullIIlllIllLLLII #12;. #12;DESIGN WINDOWS FOR A He COOLED FUSION REACTOR '....."[',":-,_.30 B93 _I_TFIII_31_ONOF THIS DO_.JMENT IS LJNLIMITED 0 S 1" I #12;,l° Design Windows for a He Cooled A design window concept is developed for a He-cooled of a helium cooled reactor are: fusion reactor blanket

Harilal, S. S.

113

Fusion component design for the moving-ring field-reversed mirror reactor  

DOE Green Energy (OSTI)

This partial report on the reactor design contains sections on the following: (1) burner section magnet system design, (2) plasma ring energy recovery, (3) vacuum system, (4) cryogenic system, (5) tritium flows and inventories, and (6) reactor design and layout. (MOW)

Carlson, G.A.

1981-01-28T23:59:59.000Z

114

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein  

E-Print Network (OSTI)

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein Argonne National Laboratory 9700 South Cass Avenue, Argonne, IL 60439 EQUATIONDERIVATION ABSTRACT A design window concept by this design window concept. INTRODUCTION Helium is an attractive coolant for both fusion and fission reactors

Harilal, S. S.

115

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

116

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect

In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

1998-05-01T23:59:59.000Z

117

Certification Report  

Science Conference Proceedings (OSTI)

Certification Report The certification of Conventional and Liberty Link™ (T25) Corn Leaf DNA Reference Materials Certified Reference Materials AOCS 0306-C and AOCS 0306-H G. Clapper and R. Ca

118

Certification Report  

Science Conference Proceedings (OSTI)

Certified Reference Materials AOCS 1206-A and AOCS 1206-B Report of the certification process for Conventional and Roundup Ready

119

Small Modular Fast Reactor Design Description Joint Effort  

NLE Websites -- All DOE Office Websites (Extended Search)

July 1, 2005 ANL-SMFR-1 July 1, 2005 ANL-SMFR-1 Small Modular Fast Reactor Design Description Joint Effort by Argonne National Laboratory (ANL) Commissariat a l'Energie Atomique (CEA) and Japan Nuclear Cycle Development Institute (JNC) Project Leaders Y. I. Chang and C. Grandy, ANL P. Lo Pinto, CEA M. Konomura, JNC Technical Contributors ANL: J. Cahalan, F. Dunn, M. Farmer, S. Kamal, L. Krajtl, A. Moisseytsev, Y. Momozaki, J. Sienicki, Y. Park, Y. Tang, C. Reed, C. Tzanos, S. Wiedmeyer, and W. Yang CEA: P. Allegre, J. Astegiano, F. Baque, L. Cachon, M. S. Chenaud, J-L Courouau, Ph. Dufour, J. C. Klein, C. Latge, C. Thevenot, and F. Varaine JNC: M. Ando, Y. Chikazawa, M. Nagamura, Y. Okano, Y. Sakamoto,

120

Improved Design of Nuclear Reactor Control System | U.S. DOE...  

Office of Science (SC) Website

Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Spinoff Applications Spinoff Archives...

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


121

ENERGY STAR Industrial Plant Certification: Professional Engineers...  

NLE Websites -- All DOE Office Websites (Extended Search)

Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program...

122

Nuclear Power 2010 Program Lessons Learned Report on the Combined Construction and Operating License/Design Certification Demonstration Projects  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Power 2010 Program Combined Construction and Operating License & Design Certification Demonstration Projects Lessons Learned Report August 30, 2012 Prepared by Longenecker and Associates DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not

123

Development of Applicable Benchmark Experiments for (Th,Pu)O2 Power Reactor Designs Using TSUNAMI Analysis.  

E-Print Network (OSTI)

?? When simulating reactor physics experiments, uncertainties in nuclear data result in a bias between simulated and experimental values. For new reactor designs or for… (more)

Langton, Stephanie E

2013-01-01T23:59:59.000Z

124

DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond  

SciTech Connect

An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

Pan, Paul Y [Los Alamos National Laboratory

2010-12-10T23:59:59.000Z

125

Why Nuclear Energy? - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

126

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

127

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

128

Design, Analysis and Optimization of the Power Conversion System for the Modular Pebble Bed Reactor System  

E-Print Network (OSTI)

technology and complies with all current codes and standards. Using the initial reference design, limiting. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion reactor design requirements...........................................39 2.5 Overall development path

129

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

130

An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor  

SciTech Connect

The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

Haihua Zhao; Hongbin Zhang

2007-11-01T23:59:59.000Z

131

LIBRA-A light ion beam fusion reactor conceptual design  

Science Conference Proceedings (OSTI)

The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4-MJ-shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbide tubes with Li/sub 17/Pb/sub 83/ flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is a right circular cylinder, 8.7 meters in diameter. The target chamber is ''self-pumped'' by the target explosion generated overpressure into a surge tank partially filled with liquid that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequency without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of LIBRA is estimated to be $2200/kWe. 12 refs., 9 figs., 1 tab.

Moses, G.A.; Kulcinski, G.L.; Bruggink, D.; Engelstad, R.; Lovell, E.; MacFarlane, J.; Musicki, Z.; Peterson, R.; Sawan, M.; Sviatoslavsky, I.

1988-01-01T23:59:59.000Z

132

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power  

Science Conference Proceedings (OSTI)

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.

John Darrell Bess

2008-06-01T23:59:59.000Z

133

TerraPower Traveling Wave Reactor: Design and Development Status...  

NLE Websites -- All DOE Office Websites (Extended Search)

Aug 28 2013 09:00 AM - 10:00 AM Pat Schweiger, TerraPower, LLC, Bellevue, Washington Reactor and Nuclear Systems Division Seminar ORNL Conference Center (Bldg. 5200), TN Rm...

134

The TITAN Reversed-Field Pinch Reactor: Design-point determination and parametric studies  

SciTech Connect

The multi-institutional TITAN study has examined the physics, technology, safety, and economics issues associated with the operation of a Reversed-Field Pinch (RFP) magnetic fusion reactor at high power density. A comprehensive system and trade study have been conducted as an integral and ongoing part of the reactor assessment. Attractive design points emerging from these parametric studies are subjected to more detailed analysis and design integration, the results of which are used to refine the parametric systems model. The design points and tradeoffs for two TITAN/RFP reactor embodiments are discussed. 14 refs.

Miller, R.L.

1987-01-01T23:59:59.000Z

135

Safety aspects of the US advanced LMR (liquid metal reactor) design  

SciTech Connect

The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs.

Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S. (Argonne National Lab., IL (USA); General Electric Co., San Jose, CA (USA); Argonne National Lab., IL (USA); USDOE Assistant Secretary for Nuclear Energy, Washington, DC (USA))

1989-01-01T23:59:59.000Z

136

Production of Advanced Biofuels via Liquefaction - Hydrothermal Liquefaction Reactor Design: April 5, 2013  

SciTech Connect

This report provides detailed reactor designs and capital costs, and operating cost estimates for the hydrothermal liquefaction reactor system, used for biomass-to-biofuels conversion, under development at Pacific Northwest National Laboratory. Five cases were developed and the costs associated with all cases ranged from $22 MM/year - $47 MM/year.

Knorr, D.; Lukas, J.; Schoen, P.

2013-11-01T23:59:59.000Z

137

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 2. MATERIALS TESTING REACTOR DESIGN AND OPERATION  

SciTech Connect

The reactor components, building, control system and circuitry, and experimental and handling facilities are described and discussed, together with operation, shutdown, tank work and supplemental facilities. Training questions and answers are included. (D.C.W.)

1963-10-01T23:59:59.000Z

138

DESIGN AND HAZARDS SUMMARY REPORT, BOILING REACTOR EXPERIMENT V (BORAX V)  

SciTech Connect

Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)

1961-05-01T23:59:59.000Z

139

Improved Design of Nuclear Reactor Control System | U.S. DOE Office of  

Office of Science (SC) Website

Improved Design of Nuclear Reactor Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Spinoff Applications Spinoff Archives SBIR/STTR Applications of Nuclear Science and Technology Funding Opportunities Nuclear Science Advisory Committee (NSAC) News & Resources Contact Information Nuclear Physics U.S. Department of Energy SC-26/Germantown Building 1000 Independence Ave., SW Washington, DC 20585 P: (301) 903-3613 F: (301) 903-3833 E: sc.np@science.doe.gov More Information » Spinoff Archives Improved Design of Nuclear Reactor Control System Print Text Size: A A A RSS Feeds FeedbackShare Page Application/instrumentation: Improved Design of Nuclear Reactor Control System Developed at: Oak Ridge National Laboratory, Holifield Radioactive Ion Beam Facility (HRIBF)

140

Advanced Reactor Design for Integrated WGS/Pre-combustion CO2...  

NLE Websites -- All DOE Office Websites (Extended Search)

Advanced Reactor Design for Integrated WGSPre-combustion CO2 Capture TDA Research, Inc. Project Number: FE0012048 Project Description The purpose is to develop a new high-hydrogen...

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


141

The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design  

SciTech Connect

The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance oflthe industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications. (FI)

Moses, D.L.; McKnight, R.D.

1987-01-01T23:59:59.000Z

142

Power conversion system design for supercritical carbon dioxide cooled indirect cycle nuclear reactors  

E-Print Network (OSTI)

The supercritical carbon dioxide (S-CO?) cycle is a promising advanced power conversion cycle which couples nicely to many Generation IV nuclear reactors. This work investigates the power conversion system design and ...

Gibbs, Jonathan Paul

2008-01-01T23:59:59.000Z

143

Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor  

E-Print Network (OSTI)

Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...

Bean, Malcolm K.

2011-08-01T23:59:59.000Z

144

Optimal design of a high pressure organometallic chemical vapor deposition reactor  

Science Conference Proceedings (OSTI)

A team composed of material scientists, physicists, and applied mathematicians have used computer simulations as a fundamental design tool in developing a new prototype High Pressure Organometallic Chemical Vapor Deposition (HPOMCVD) reactor for use ...

K. J. Bachmann; H. T. Banks; C. Höpfner; G. M. Kepler; S. Lesure; S. D. Mccall; J. S. Scroggs

1999-04-01T23:59:59.000Z

145

Design of Slurry Bubble Column Reactors: Novel Technique for Optimum Catalyst Size Selection  

NLE Websites -- All DOE Office Websites (Extended Search)

Slurry Bubble Column Reactors: Novel Technique Slurry Bubble Column Reactors: Novel Technique for Optimum Catalyst Size Selection Opportunity The Department of Energy's National Energy Technology Laboratory (NETL) is seeking licensing partners interested in implementing United States Patent Number 7,619,011 entitled "Design of Slurry Bubble Column Reactors: Novel Technique for Optimum Catalyst Size Selection." Disclosed in this patent is a method to determine the optimum catalyst particle size for application in a fluidized bed reactor, such as a slurry bubble column reactor (SBCR), to convert synthesis gas into liquid fuels. The reactor can be gas-solid, liquid- solid, or gas-liquid-solid. The method considers the complete granular temperature balance based on the kinetic theory of

146

CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide  

Science Conference Proceedings (OSTI)

CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels.

Thompson, W.I.

1980-10-01T23:59:59.000Z

147

Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report  

Science Conference Proceedings (OSTI)

This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

1992-03-01T23:59:59.000Z

148

Design, Fabrication, and Certification of Advanced Modular PV Power Systems Final Technical Progress Report  

DOE Green Energy (OSTI)

This report describes the overall accomplishments and benefits of Solar Electric Specialties Co. (SES) under this Photovoltaic Manufacturing Technology (PVMaT) subcontract. SES addressed design issues related to their modular autonomous PV power supply (MAPPS) and a mobile photogenset. MAPPS investigations included gel-cell batteries mounted horizontally; redesign of the SES power supply; modified battery enclosure for increased safety and reduced cost; programmable, interactive battery charge controllers; and UL and FM listings. The photogenset systems incorporate generators, battery storage, and PV panels for a mobile power supply. The unit includes automatic oil-change systems for the propane generators, collapsible array mounts for the PV enclosure, and internal stowage of the arrays. Standardizing the products resulted in product lines of MAPPS and Photogensets that can be produced more economically and with shorter lead times, while increasing product quality and reliability. Product assembly and quality control have also been improved and streamlined with the development of standardized assembly processes and QC testing procedures. SES offers the UL-listed MAPPS at about the same price as its previous non-standardized, unlisted products.

Lambarski, T.; Minyard, G. (Solar Electric Specialties Co., Willits, California)

1998-10-06T23:59:59.000Z

149

Design Concept and Application of Small Nuclear Power Reactor  

Science Conference Proceedings (OSTI)

The outline of the recent design concepts and those features of the small nuclear power rector are described, including specifications, present design status, application and so on.

Minato, Akio [CRIEPI, Central Research Institute of Electric Power Industry, Tokyo (Japan); Sekimoto, Hiroshi [Center for Research into Innovative Nuclear Energy Systems (CRINES) Tokyo Institute of Technology 2-12-1, Ookayama, Meguro-ku, Tokyo, 152-8550 (Japan)

2009-03-31T23:59:59.000Z

150

Certification Report  

Science Conference Proceedings (OSTI)

Certified Reference Material AOCS 0806-A AOCS 0806-B AOCS 0806-C AOCS 0806-D Report of the certification process for Conventional and EH92-527-1 Potato Potato Certified Reference Materials G. Clapper a

151

Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor  

E-Print Network (OSTI)

Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

Ellis, Tyler Shawn

2009-01-01T23:59:59.000Z

152

A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.  

SciTech Connect

many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

2008-04-23T23:59:59.000Z

153

Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor  

Science Conference Proceedings (OSTI)

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

B. Boer; A. M. Ougouag

2010-09-01T23:59:59.000Z

154

Strategic planning for LEED certification  

E-Print Network (OSTI)

Intel Corporation has recently implemented a "green building" policy, which states that Intel will design all new facilities to achieve a minimum LEED-Silver certification. LEED (Leadership in Energy and Environmental ...

Capuzzi, Angelo Michael

2010-01-01T23:59:59.000Z

155

Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors  

SciTech Connect

This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

Orvis, D.D.; Raabe, P.H.

1980-01-01T23:59:59.000Z

156

DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS  

SciTech Connect

The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

2003-11-12T23:59:59.000Z

157

Design of a 2400MW liquid-salt cooled flexible conversion ratio reactor  

E-Print Network (OSTI)

A 2400MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCI2 (30%-20%-50%) as coolant. The reference design uses a wire-wrapped, hex lattice core, and is ...

Petroski, Robert C

2008-01-01T23:59:59.000Z

158

Designing nanostructured heterogeneous catalysts to exploit pulsing in gas-liquid packed bed reactors  

E-Print Network (OSTI)

41 Designing nanostructured heterogeneous catalysts to exploit pulsing in gas-liquid packed bed nanostructured catalysts for gas-liquid reactions, which have a system of macro pores designed to take advantage in volume of gas-liquid packed bed reactors (a.k.a. "trickle" beds) by an order of magnitude or more because

McCready, Mark J.

159

Design and Manufacture of the Storage Cask for the Old Reactor Internals  

SciTech Connect

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of upper reactor internals (UCI) and lower reactor internals (LCI) of the pressurized water reactor in Shikoku Electric Power Company's Ikata Unit No.1 by 'the all-in-one-piece extraction method' introduced in the document of [ICONE14-89233]. In the pressurized water reactor (PWR) plant, the UCI are usually removed from the reactor vessel (RV) independently and reinstalled into the RV again every refueling outage. The LCI are independently able to be removed from the RV and reinstalled again during in-service inspection, too. In the boiling water reactor (BWR) plant, there are several cases of replacing BWR shrouds by cutting small and containing in a container. But no replacement of all reactor internals (CI) for the PWR, in one piece without splitting or cutting, has been reported. The purpose of this paper is to introduce the key points about the design and manufacture of the storage cask for old reactor internals in the replacement work by 'the all-in-one-piece extraction method'. (author)

Yasuhiro Tomiita [Mitsubishi Heavy Industries, Ltd. (Japan)

2006-07-01T23:59:59.000Z

160

Design studies of the Moderated Thermonic Heat Pipe Reactor (MOHTR) concept  

DOE Green Energy (OSTI)

Design studies, based primarily on neutronics analysis, have been conducted on a thermionic reactor concept that uses a combined beryllium and zirconium hydride moderator to facilitate the incorporation of heat pipe cooling into compact thermionic fuel element (TFE) based designs useful in the tens of kilowatts electrical power regime. The goal of the design approach is to achieve a single point failure free system with technologies such as TFEs, high-temperature heat pipes, and ZrH moderation, which have extensive test data bases and have been shown to be capable of long lifetimes. Beryllium is used to thermally couple redundant heat pipes to TFEs and ZrH is added to reduce critical size. Neutronic analysis undertaken to investigate this design approach shows that greater reactivity can be achieved for a given geometry with a combination of the two moderator materials than with ZrH alone and that the combined moderator is much less sensitive to hydrogen loss than more traditional ZrH-moderated thermionic reactor designs. These and other analytical approaches have demonstrated the credibility of a heat pipe cooled thermionic reactor concept that has a reactor height and diameter of 60 cm and a reactor mass of 400 kg for 30-kWe power output. 14 refs., 8 figs.

Ranken, W.A.; Turner, J.A.

1991-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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
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161

Neutronic design studies for an unattended, low power reactor  

DOE Green Energy (OSTI)

The Los Alamos National Laboratory is involved in the design and demonstrations of a small, long-lived nuclear heat and electric power source for potential applications at remote sites where alternate fossil energy systems would not be cost effective. This paper describes the neutronic design analysis that was performed to arrive at two conceptual designs, one using thermoelectric conversion, the other using an organic Rankine cycle. To meet the design objectives and constraints a number of scoping and optimization studies were carried out. The results of calculations of control worths, temperature coefficients of reactivity and fuel depletion effects are reported.

Palmer, R.G.; Durkee, J.W. Jr.

1986-01-01T23:59:59.000Z

162

Design of Materials Testing Capsule in PULSTAR Reactor for High ...  

Science Conference Proceedings (OSTI)

We report the main features of the irradiation capsule design and results from ... Synchrotron Radiation Study of Hydride Reorientation in Zircaloy under In Situ ...

163

Manhattan Project: Final Reactor Design and X-10, 1942-1943  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge FINAL REACTOR DESIGN AND X-10 Schematic of the X-10 Graphite Reactor, Oak Ridge FINAL REACTOR DESIGN AND X-10 (Met Lab and Oak Ridge [Clinton], 1942-1943) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 Before any plutonium could be chemically separated from uranium for a bomb, however, that uranium would first have to be irradiated in a production pile. CP-1 had been a success as a scientific experiment, but the pile was built on such a small scale that recovering any significant amounts of plutonium from it was impractical. In the fall of 1942, scientists of the Met Lab had decided to build a second Fermi pile at Argonne as soon as his experiments on the first were completed and to proceed with the "Mae West" design for a helium-cooled production pile as well. When DuPont engineers assessed the Met Lab's plans in the late fall, they agreed that helium should be given first priority. They placed heavy water second and urged an all-out effort to produce more of this highly effective moderator. Bismuth and water were ranked third and fourth in DuPont's analysis. Priorities began to change when Enrico Fermi's CP-1 calculations demonstrated a higher value for the neutron reproduction factor k (for a theoretical reactor of infinite size) than anyone had anticipated. Met Lab scientists concluded that a water-cooled pile was now feasible. Crawford Greenewalt, head of the DuPont effort, continued, however, to support helium cooling.

164

Material Control and Accounting Design Considerations for High-Temperature Gas Reactors  

Science Conference Proceedings (OSTI)

The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in this report, namely the development of guidance to support the consideration of MC&A in the design of both pebble-bed and prismatic-fueled HTGRs. The objective is to identify and incorporate design features into the facility design that will cost effectively aid in making MC&A more effective and efficient, with minimum impact on operations. The theft of nuclear material is addressed through both MC&A and physical protection, while the threat of sabotage is addressed principally through physical protection.

Trond Bjornard; John Hockert

2011-08-01T23:59:59.000Z

165

Implications of high efficiency power cycles for fusion reactor design  

SciTech Connect

The implications of the High Efficiency Power Cycle for fusion reactors are examined. The proposed cycle converts most all of the high grade CTR heat input to electricity. A low grade thermal input (T approximately 100$sup 0$C) is also required, and this can be supplied at low cost geothermal energy at many locations in the U. S. Approximately 3 KW of low grade heat is required per KW of electrical output. The thermodynamics and process features of the proposed cycle are discussed. Its advantages for CTR's are that low Q machines (e.g. driven Tokamaks, mirrors) can operate with a high (approximately 80 percent) conversion of CTR fusion energy to electricity, where with conventional power cycles no plant output could be achieved with such low Q operation. (auth)

Powell, J.R.; Usher, J.; Salzano, F.J.

1975-01-01T23:59:59.000Z

166

Compliance Certification Application  

NLE Websites -- All DOE Office Websites (Extended Search)

40 CFR Part 191 40 CFR Part 191 Compliance Certification Application for the Waste Isolation Pilot Plant Appendix PIC United States Department of Energy Waste Isolation Pilot Plant Carlsbad Area Office Carlsbad, New Mexico Passive Institutional Controls Conceptual Design Report PASSIVE INSTITUTIONAL CONTROLS CONCEPTUAL DESIGN REPORT REVISION 0 MAY 14, 1996 1 C PASSIVE INSTITUTIONAL CONTROLS DESIGN REPORT REVISION 0 PASSIVE INSTITUTIONAL CONTROLS CONCEPTUAL DESIGN REPORT MAY 1 4 . 1 9 9 6 TABLE OF CONTENTS INTRODUCTION . . . . . . . . . . . . . . . . . . . . 1 A . Purpose . . . . . . . . . . . . . . . . . . . 1 B . Scope . . . . . . . . . . . . . . . . . . . . . 3 C . Background . . . . . . . . . . . . . . . . . . . 7 I1 . SITE DESCRIPTION . . . . . . . . . . . . . . . . . 11 I11 . DESIGN REQUIREMENTS/CRITERIA . . . . . . . . . . 17 IV . MESSAGES . . . . . . . . . . . . . . . . . .

167

SUMMARY OF REACTOR DESIGN INFORMATION FROM THREE YEARS' OPERATION OF A SMALL PWR  

SciTech Connect

Reactor design information obtained from 3 years' operation of a small pressurized-water reactor, the SM-1 (formerly APPR-l), is presented and discussed. The SM-1 reactor, designed to produce 10 Mw(t) power, employs fully enriched uranium fuel in the form of UO/sub 2/ dispersed in stainless-steel fuel plates. The reactor is cooled by water at 1200 psia and mean temperature of 44) deg F. Core-physics measurements were performed of temperature coefficient, pressure coefficient, rod calibration, stuck rod position, and transient xenon as a function of core burn-out. Core burn-out characteristics were compared with few- group calculations, and reasonable agreement was obtained. Thermal-heat-balance data were obtained on the reactor core. The temperature pattern in the nominal and hot channels under operating conditions was calculated. These calculations indicated that certain of the fuel channels operated in the nucleate boiling regime. Examination of one of the fuel channels suspected of nucleate boiling indicated no adverse effects. The system response to load perturbations and during pump coast-down was measured utilizing plant instrumentation. This response was compared with analytical predictions using a lumped kinetic model, and reasonable agreement was found. Both neutron and gamma traverses were made through the primary shield during reactor operation. Gamma traverses were also made through the primary shield as a function of time after reactor shutdown. Conventional shielding calculational methods are found to give agreement with experiment sufficient for design purposes. An absolute ionization chamber was employed to measure N/sup 16/ activity in the reactor coolant. These measurements were compared with N/sup 16/ calculated from the (n,p) reaction on O/ sup 16/. (auth)

Gallagher, J.G.

1960-09-01T23:59:59.000Z

168

A Review of the Containment Building Design for the Advanced Reactor  

SciTech Connect

A pilot plant is being designed to prove and validate the technical merits and capabilities of the System-Integrated Modular Advanced Reactor(SMART) technology. The first phase of architect/engineering services is being in progress to obtain the construction permit for the pilot plant. During this first phase, the Safe Guard Vessel that surrounds the reactor vessel was eliminated and its function incorporated into the containment building structure. Further investigation and review were performed to optimize the Reactor Containment Building structure and the layout inside to ensure all design criteria and concepts required by the SMART technology were met. This paper describes the review and the design of the Reactor Containment Building structure for the pilot plant considering the requirements of the original SMART design. The results of this review show that the cylindrical reinforced concrete containment was selected from the various types of the containment buildings and will be used to demonstrate the performance of the original SMART reactor. (authors)

Lee, Joon-Ho; Park, Mun-Baek; Yun, Soon-Chul [Korea Power Engineering Company, Inc., 360-9 Mabuk-Ri, Gusong-Eup, Yongin-Si, Kyonggi-Do, 449-713 (Korea, Republic of)

2004-07-01T23:59:59.000Z

169

Modularity in design of the MIT Pebble Bed Reactor  

E-Print Network (OSTI)

The future of new nuclear power plant construction will depend in large part on the ability of designers to reduce capital, operations, and maintenance costs. One of the methods proposed, is to enhance the modularity of ...

Berte, Marc Vincent, 1977-

2004-01-01T23:59:59.000Z

170

International Thermonuclear Experimental Reactor (ITER) neutral beam design  

SciTech Connect

This report discusses the following topics on ITER neutral beam design: ion dump; neutralizer and module gas flow analysis; vacuum system; cryogenic system; maintainability; power distribution; and system cost.

Myers, T.J.; Brook, J.W.; Spampinato, P.T.; Mueller, J.P.; Luzzi, T.E.; Sedgley, D.W. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)

1990-10-01T23:59:59.000Z

171

Benchmark analysis for the design of piping systems in advanced reactors  

SciTech Connect

To satisfy the need for the verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boding water reactor standard design, three piping benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set. A summary description of each problem and some sample results are included.

Bezler, P.; DeGrassi, G.; Braverman, J. (Brookhaven National Lab., Upton, NY (United States)); Shounien Hou (Nuclear Regulatory Commission, Washington, DC (United States))

1993-01-01T23:59:59.000Z

172

Benchmark analysis for the design of piping systems in advanced reactors  

Science Conference Proceedings (OSTI)

To satisfy the need for the verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boding water reactor standard design, three piping benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set. A summary description of each problem and some sample results are included.

Bezler, P.; DeGrassi, G.; Braverman, J. [Brookhaven National Lab., Upton, NY (United States); Shounien Hou [Nuclear Regulatory Commission, Washington, DC (United States)

1993-03-01T23:59:59.000Z

173

Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor  

Science Conference Proceedings (OSTI)

A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

Ishii, M.; Xu, Y.; Revankar, S.T. [Purdue University, West Lafayette, IN 47907 (United States)

2002-07-01T23:59:59.000Z

174

DESIGN AND FEASIBILITY STUDY OF A PEBBLE BED REACTOR-STEAM POWER PLANT  

SciTech Connect

Originally issued as S and P 1963A, Parts I and II. A design and feasibility study of a pebble bed reactorsteam power plant is presented, The reactor design which evolved from this study is a 125 Mwe heliumcooled two-region thermal breeder, operating on the uranium-thorium cycle, in which all core structural materials are graphite. Fuel is in the form of unclad spherical elements of graphite, containing fissile and fertile material. The primary loop consists of the reactor plus three steam generators and blowers in parallel. Nuclear characteristics, costs, etc., are given. (W.D.M.)

1958-05-01T23:59:59.000Z

175

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor  

SciTech Connect

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

2010-09-01T23:59:59.000Z

176

Design strategies for optically-accessible, high-temperature, high-pressure reactor  

Science Conference Proceedings (OSTI)

The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

2000-02-01T23:59:59.000Z

177

Design Strategies for Optically-Accessible, High-Temperature, High-Pressure Reactor  

SciTech Connect

The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

2000-02-01T23:59:59.000Z

178

An autonomous long-term fast reactor system and the principal design limitations of the concept  

E-Print Network (OSTI)

The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National Laboratory. As a result of the computational analysis performed in this work, the ALM-FR design provides for the possibility of continuous operation during about 40 years on one fuel loading containing mixture of depleted uranium with plutonium and higher actinides. All reactor physics characteristics of the ALM-FR are kept within technological limits ensuring safety of ultra-long autonomous operation. The results obtained provide for identification of physical features of the ALM-FR that significantly influence flexibility of the design and its applications. The special emphasis is given to existing limitations on the utilization of higher actinides as a fuel component.

Tsvetkova, Galina Valeryevna

2003-12-01T23:59:59.000Z

179

Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies  

SciTech Connect

The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %dk/k to 0.0004 %dk/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.

John D. Bess; Margaret A. Marshall

2013-02-01T23:59:59.000Z

180

Options Study Documenting the Fast Reactor Fuels Innovative Design Activity  

Science Conference Proceedings (OSTI)

This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

Jon Carmack; Kemal Pasamehmetoglu

2010-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" from the National Library of EnergyBeta (NLEBeta).
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181

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

2006-10-01T23:59:59.000Z

182

Status report on the conceptual design of a commercial tokamak hybrid reactor (CTHR)  

SciTech Connect

A preliminary conceptual design is presented for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) plants. The study has been made in sufficient depth to indicate no insurmountable technical problems exist and has provided a basis for valid cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources.

1979-09-01T23:59:59.000Z

183

Reference (Axially Graded) Low Enriched Uranium Fuel Design for the High Flux Isotope Reactor (HFIR)  

Science Conference Proceedings (OSTI)

During the past five years, staff at the Oak Ridge National Laboratory (ORNL) have studied the issue of whether the HFIR could be converted to low enriched uranium (LEU) fuel without degrading the performance of the reactor. Using state-of-the-art reactor physics methods and behind-the-state-of-the-art thermal hydraulics methods, the staff have developed fuel plate designs (HFIR uses two types of fuel plates) that are believed to meet physics and thermal hydraulic criteria provided the reactor power is increased from 85 to 100 MW. The paper will present a defense of the results by explaining the design and validation process. A discussion of the requirements for showing applicability of analyses to approval for loading the fuel to HFIR lead test core irradiation currently scheduled for 2016 will be provided. Finally, the potential benefits of upgrading thermal hydraulics methods will be discussed.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

184

CONCEPTUAL DESIGN AND ECONOMIC EVALUATION OF A STEAM-COOLED FAST BREEDER REACTOR  

SciTech Connect

A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estimated, based on AEC ground rules for privately financed plants and utilities. This cost may go down to 5.7 mills/kwh by 1975. For the 40 MW/sub e/ plant corresponding energy costs are 19.5 and 13.7 mills/kwh, r -spectively. The R&D program required for this reactor concept is estimated at million with an additional million for improvements leading to the 1975 reactor. Investigation of the operational and safety aspects of the reactor indicated that satisfactory procedures can be used for startup, shutdown, and emergency cooling of the reactor. An increase in reactivity upon flooding can be prevented by incorprating small amounts of high resonance absorption material in the core. Preliminary calculations indicate a substantial increase in reactivity upon loss of coolant for the 300 MW/sub e/ PuO/sub 2/ fueled reactor. To obtain designs with satisfactory voiding characteristics it may be necessary to provide high neutron leakage as ib a low L/D core or smaller volume core. Acceptable voiding characteristics appear possible with a Pu fueled 40 MWe reactor cooled with H/O steam, a Pu fueled 300MW/sub w/,reactor cooled with D/O steam, and a 300 MW/sub e/ U/sup 233/-Th fueled 300 MW/sub e/ breeder reactor cooled with H/sub 2/O steam. (auth)

Sofer, G.; Hankel, R.; Goldstein, L.; Birman, G.

1961-11-15T23:59:59.000Z

185

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

SciTech Connect

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

186

Spring design for use in the core of a nuclear reactor  

DOE Patents (OSTI)

A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

Willard, Jr., H. James (Bethel Park, PA)

1993-01-01T23:59:59.000Z

187

DESIGN CRITERIA FOR HIGH TEMPERATURE LATTICE TEST REACTOR PROJECT CAH-100  

SciTech Connect

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)

Ballard, D.L.; Brown, W.W.; Harrison, C.W.; Heineman, R.E.; Henry, H.L.; Jeffs, T.W.; Morrow, G.W.; Russell, J.T.; Waite, J.K.

1963-05-24T23:59:59.000Z

188

DESIGN STUDY OF SMALL BOILING REACTORS FOR POWER AND HEAT PRODUCTION  

SciTech Connect

A design study has been made of a small "Package" nuclear power plant for the production of electric power and heat in remotely located, inaccessible areas devoid of natural fuels. The design utilizes a horizontal boiling reactor as a steam generator consistent with safe and simple equipment and a minimum building height. A reactor design of 51/2 Mw capacity, with a combined net electric power output of 750 kw and a heat plant output of 4500 kw, was studied in detail. Tertative cost estimates are presented on the basis of this combination. General comparisons have been made between different systems designed for either independent or combined production of 425 kw net electric power and 2500 kw available heat. (auth)

Treshow, M.

1954-11-01T23:59:59.000Z

189

Safety goals and functional performance criteria. [Advanced Reactor Design  

DOE Green Energy (OSTI)

This report discusses a possible approach to the development of functional performance criteria to be applied to evolutionary LWR designs. Key safety functions are first identified; then, criteria are drawn up for each individual function, based on the premise that no single function's projected unreliability should be allowed to exhaust the safety goal frequencies. In the area of core damage prevention, functional criteria are cast in terms of necessary levels of redundancy and diversity of critical equipment. In the area of core damage mitigation (containment), functional performance criteria are cast with the aim of mitigating post-core-melt phenomena with sufficient assurance to eliminate major uncertainties in containment performance. 9 refs.

Youngblood, R.W.; Pratt, W.T. (Brookhaven National Lab., Upton, NY (USA)); Hardin, W.B. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of Regulatory Applications)

1990-01-01T23:59:59.000Z

190

Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)  

DOE Green Energy (OSTI)

A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the equivalent temperature of heat delivered to either the power conversion system or a hydrogen production plant. Using a comparative cost analysis, the construction costs per unit output are projected to be 50-55% of the costs for modular gas-cooled or sodium-cooled reactor systems. This is primarily a consequence of substantially larger power output and higher conversion efficiency for the AHTR. The AHTR has a number of unique technical challenges in meeting the NGNP requirements; however, it appears to offer advantages over high-temperature helium-cooled reactors and provides an alternative development path to achieve the NGNP requirements. Primary challenges include optimizing the core design for improved response to transients, designing an internal blanket to thermally protect the reactor vessel, and engineering solutions to high-temperature refueling and maintenance.

Ingersoll, D.T.

2004-07-29T23:59:59.000Z

191

Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask  

Science Conference Proceedings (OSTI)

Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

Pope, R.B.; Diggs, J.M. (eds.)

1982-04-01T23:59:59.000Z

192

THE CNSG II--A CONCEPTUAL MERCHANT SHIP NUCLEAR REACTOR DESIGN  

SciTech Connect

The Consolidated Nuclear Steam Generator H consists of a pressurized water reactor, a steam generator, and a pressurizer combined in a sirgle pressure vessel. The design of the 66000 shaft horsepower system is presented, together with basic plant irformation for designs of 15000, 22000, and 30000 shaft horsepower. The economics, safety characteristics, and operational procedures of the plant are also discussed. (D.C.W.)

1963-09-01T23:59:59.000Z

193

Presentation of NNSA Certificates to the Argonne RAP Team  

NLE Websites -- All DOE Office Websites (Extended Search)

Presentation of NNSA Certificates to Argonne's RAP Team Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Presentation of NNSA Certificates to Argonne's RAP Team Group photo taken during the presentation of the Letter of Appreciation Pictured in photo left to right: R. Klann (NE), F. Moore (NE), S.

194

PRISM: An innovative liquid metal fast breeder reactor  

SciTech Connect

This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design.

Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

1986-01-01T23:59:59.000Z

195

A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel  

Science Conference Proceedings (OSTI)

An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

Samuel E. Bays

2011-02-01T23:59:59.000Z

196

A helium-cooled blanket design of the low aspect ratio reactor  

Science Conference Proceedings (OSTI)

An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

Wong, C.P.; Baxi, C.B.; Reis, E.E. [General Atomics, San Diego, CA (United States); Cerbone, R.; Cheng, E.T. [TSI Research, Solana Beach, CA (United States)

1998-03-01T23:59:59.000Z

197

Mechanical and thermal design of the Cascade reactor  

Science Conference Proceedings (OSTI)

We present an improved Cascade fusion reaction chamber that is optimized with respect to chamber radius, wall thickness, and pebble blanket outlet temperature. We show results of a parameter study where we varied chamber radius from 3 to 6 m, wall thickness from 15 to 80 mm, and blanket outlet temperature from 900 to 1400 K. Based on these studies, we achieved an optimized chamber with 50% the volume of the original design and 60% of its blanket. Chamber radius is only 4.4 m and its half length is only 5.9 m, decreased from the original 5-m radius and 8-m half-length. In our optimization method, we calculate both thermal and mechanical stresses resulting from x-ray, fusion-pellet-debris, and neutron-generated momentum, pressure from ablated material, centrifugal action, vacuum inside the chamber, and gravity. We add the mechanical stresses to thermal stresses and keep the total less than the yield stress. Further, we require that fluctuations in these stresses be less than that which would produce creep-fatigue failure within the chamber 30-year lifetime.

Pitts, J.H.

1983-01-01T23:59:59.000Z

198

Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors  

SciTech Connect

The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO{sub 2}-UO{sub 2}) fuel and compare those designs with more conventional UO{sub 2} designs.The fuel cycle analyses indicate that ThO{sub 2}-UO{sub 2} fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO{sub 2} fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO{sub 2}-UO{sub 2} fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO{sub 2} fuel used in light water reactors.

Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A. [Framatome ANP, Inc. (France)

2004-07-15T23:59:59.000Z

199

Core and Refueling Design Studies for the Advanced High Temperature Reactor  

SciTech Connect

The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure drop across the core was restricted to no more than 1.5 atm during normal operation to minimize the upward force on the core. Also, the flow velocity in the core was restricted to 3 m/s to minimize erosion of the fuel plates. Section 3.1.1 of this report discusses the design restrictions in more detail.

Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

2011-09-01T23:59:59.000Z

200

Containment Versus Confinement for High-Temperature Gas Reactors: Regulatory, Design Basis, Siting, and Cost/Economic Considerations  

Science Conference Proceedings (OSTI)

This report provides the results of an investigation pertaining to the use of the confinement that has been proposed for the high temperature and very high temperature gas reactors (HTGR, VHTR). No comprehensive study of this question has been published since 1985. All power reactor designs to go into commercial service in the United States were light water reactors (LWR), except for Fort St. Vrain (FSV) and Peach Bottom Unit 1, which were steam cycle helium gas cooled reactors. All designs use a leak-ti...

2005-05-04T23:59:59.000Z

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


201

Preliminary safety calculations to improve the design of Molten Salt Fast Reactor  

SciTech Connect

Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A. [LPSC, CNRS/IN2P3, Grenoble INP, 53,rue des Martyrs, 38026 Grenoble Cedex (France)

2012-07-01T23:59:59.000Z

202

Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors  

Science Conference Proceedings (OSTI)

Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

2012-12-01T23:59:59.000Z

203

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

and A. SESONSKE. Nuclear Reactor Engineering: Third Edition.E. LEWIS. Fundamentals of Nuclear Reactor Physics. Elseviervan DAM. “Physics of nuclear reactor safety”. In: Reports on

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

204

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network (OSTI)

L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

Galvez, Cristhian

2011-01-01T23:59:59.000Z

205

DESIGN OF A TOKAMAK FUSION REACTOR FIRST WALL ARMOR AGAINST NEUTRAL BEAM IMPINGEMENT  

E-Print Network (OSTI)

Hoffman, et. a1. , "Fusion Reactor First Wall Cooling forTheir Signif- icance in Fusion Reactors," Fifth ConferenceProb- lems in Toroidal Fusion Reactors," Fifth Conference

Myers, Richard Allen

2011-01-01T23:59:59.000Z

206

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

well known from basic reactor theory, the flux distributionof a fast reactor using the perturbation theory”. In: Atomicbeam theory and are not specific to a nuclear reactor core.

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

207

PRELIMINARY DESIGN STUDY FOR A SODIUM-GRAPHITE-REACTOR IRRADIATION FACILITY  

SciTech Connect

The results of an investigation to integrate a Na/sup 24/ irradiation processing facility with an operating sodium graphite reactor are presented. An irradiation facility incorporated into a reference SGR (Hallam Nuclear Power Facility, Hallam, Nebraska) is described. Development of the facility application, preliminary design criteria and capital and operating costs are discussed. Recommendations for further development of the technology and economics of this type of irradiation facility are included. (auth)

Thompson, D.S.; Benaroya, V.

1959-01-31T23:59:59.000Z

208

A DESIGN STUDY OF A LOW POWER AQUEOUS HOMOGENEOUS BOILING REACTOR POWER PLANT  

SciTech Connect

This design study describes a reactor and associated power plant that has been designed to produce 100 kv of net electric power and 400 kv of hot water space heating at a total thermal output of 1300 kw. The fuel consists of a solution of UO/sub 2/SO/sub 4/ in light water. Power is removed from the core by boiling the fuel solution and transferring the heat to the secondary steam system by condensing primary water on the external surface of a bayonet type boiler and boiling secondary water within the tubes. Saturated steam, produced in the boiler at 225 psia (Full Power) is used to drive a turbo generator, Extraction steam from the turbine is used, at a reduced pressure, for space heating. The initial loading of the reactor is approximately 4.8 kg of U/sub 235/ and operation based on an average load factor of 80% will require fuel addition at the rate of about 580 grams per year. It may be desirable to replace the fuel in the core after a period of 5 years operation due to the accumulation of corrosion products. The reactor control is affected automatically by power demand. The major objective has been to design a reactor that is reliable and simple, requiring little if any operating personnel and routine maintenance only which can be performed by one man. The design should stress simplicity of the system, ease of erection at the site, initial transportability, reliability and ease of operation; these characteristics are then expected to result in greatly reduced effort and manpower support over a conventional system. (auth)

Mong, B.A.; Colgan, J.E.; D' Elia, R.A.; Mooradian, J.S.; Rhode, G.K.; Wood, P.M.

1955-06-01T23:59:59.000Z

209

The Treatment of Renewable Energy Certificates, Emissions Allowances, and Green Power Programs in State Renewables Portfolio Standards  

E-Print Network (OSTI)

Holt. 2004. “Design Guide for Renewable Energy Certificate2005. “Emerging Markets for Renewable Energy Certificates:Golden, CO: National Renewable Energy Laboratory. http://

Holt, Edward A.; Wiser, Ryan H.

2007-01-01T23:59:59.000Z

210

The Treatment of Renewable Energy Certificates, Emissions Allowances, and Green Power Programs in State Renewables Portfolio Standards  

E-Print Network (OSTI)

Design Guide for Renewable Energy Certificate Tracking2005. “Emerging Markets for Renewable Energy Certificates:Golden, CO: National Renewable Energy Laboratory. http://

Holt, Edward A.; Wiser, Ryan H.

2007-01-01T23:59:59.000Z

211

Experimental and design experience with passive safety features of liquid metal reactors  

SciTech Connect

Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-II with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions.

Lucoff, D.M.; Waltar, A.E. [Westinghouse Hanford Co., Richland, WA (United States); Sackett, J.I. [Argonne National Lab., Idaho Falls, ID (United States); Salvatores, M. [CEA, 75 - Paris (France); Aizawa, K. [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

1992-10-01T23:59:59.000Z

212

Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters  

Science Conference Proceedings (OSTI)

Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with â??warm boreâ?ť diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged â??spiderâ?ť design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project â??Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limitersâ?ť was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZPâ??s product development program, the amount of HTS wire employed per FCL and its cost as a percentage of the total FCL product content had not dropped substantially from an unsustainable level of more than 50% of the total cost of the FCL, nor had the availability increased (today the availability of 2G wire for commercial applications outside of specific partnerships with the leading 2G wire manufacturers is extremely limited). ZP had projected a very significant commercial potential for FCLs with higher performance and lower costs compared to the initial models built with 1G wire, which would come about from the widespread availability of low-cost, high-performance 2G HTS wire. The potential for 2G wires at greatly reduced performance-based prices compared to 1G HTS conductor held out the potential for the commercial production of FCLs at price and performance levels attractive to the utility industry. However, the price of HTS wire did not drop as expected and today the available quantities of 2G wire are limited, and the price is higher than the currently available supplies of 1G wire. The commercial option for ZP to provide a reliable and reasonably priced FCL to the utility industry is to employ conventional resistive conductor DC electromagnets to bias the FCL. Since the premise of the original funding was to stimulate the HTS wire industry and ZP concluded that copper-based magnets were more economical for the foreseeable future, DOE and ZP decided to mutually terminate the project.

Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

2011-10-31T23:59:59.000Z

213

Design Configurations for a Very High Temperature Gas-Cooled Reactor Designed to Generate Electricity and Hydrogen  

DOE Green Energy (OSTI)

The High Temperature Gas-Cooled Reactor is being envisioned that will generate not just electricity, but also hydrogen to charge up fuel cells for cars, trucks and other mobile energy uses. INL engineers studied various heat-transfer working fluids—including helium and liquid salts—in seven different configurations. In computer simulations, serial configurations diverted some energy from the heated fluid flowing to the electric plant and hydrogen production plant. In anticipation of the design, development and procurement of an advanced power conversion system for HTGR, this study was initiated to identify the major design and technology options and their tradeoffs in the evaluation of power conversion system (PCS) coupled to hydrogen plant. In this study, we investigated a number of design configurations and performed thermal hydraulic analyses using various working fluids and various conditions (Oh, 2005). This paper includes a portion of thermal hydraulic results based on a direct cycle and a parallel intermediate heat exchanger (IHX) configuration option.

Conference preceedings

2006-07-01T23:59:59.000Z

214

Standards and Certification  

Science Conference Proceedings (OSTI)

... Standards and Certification. Voluntary Voting System Guidelines (VVSG). The Help America Vote Act instructed the Election ...

2013-09-17T23:59:59.000Z

215

IBM Certification White Paper  

E-Print Network (OSTI)

IBM Certification White Paper The value of IT Certification #12;The value of IT Certification White · Respect and admiration of peers · Entrance into a community of like-minded professionals This white paper that can be evaluated in terms of ROI. Those most often #12;The value of IT Certification White paper

216

AOCS Membership Certificate  

Science Conference Proceedings (OSTI)

Certificates are 8 1/2 x 11 and are suitable for framing. AOCS Membership Certificate Membership Merchandise Membership Merchandise These two color certificates are signed by the AOCS President.  Certificates are 8 1/2 x 11 and are suitable for fr

217

A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design  

SciTech Connect

Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

2010-11-01T23:59:59.000Z

218

Analysis of Reference Design for Nuclear-Assisted Hydrogen Production at 750°C Reactor Outlet Temperature  

SciTech Connect

The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using a high-temperature gas-cooled reactor (HTGR) to provide the process heat and electricity to drive the electrolysis process. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This report describes the resulting new INL reference design coupled to two alternative HTGR power conversion systems, a Steam Rankine Cycle and a Combined Cycle (a Helium Brayton Cycle with a Steam Rankine Bottoming Cycle). Results of system analyses performed to optimize the design and to determine required plant performance and operating conditions when coupled to the two different power cycles are also presented. A 600 MWt high temperature gas reactor coupled with a Rankine steam power cycle at a thermal efficiency of 44.4% can produce 1.85 kg/s of hydrogen and 14.6 kg/s of oxygen. The same capacity reactor coupled with a combined cycle at a thermal efficiency of 42.5% can produce 1.78 kg/s of hydrogen and 14.0 kg/s of oxygen.

Michael G. McKellar; Edwin A. Harvego

2010-05-01T23:59:59.000Z

219

Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core  

SciTech Connect

A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

Sterbentz, James W

2007-05-01T23:59:59.000Z

220

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

Science Conference Proceedings (OSTI)

A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

2004-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors.  

E-Print Network (OSTI)

??Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical… (more)

Samuel, Jeffrey

2011-01-01T23:59:59.000Z

222

Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics  

SciTech Connect

A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

Adams, S.R.

1985-10-01T23:59:59.000Z

223

Progress on the conceptual design of a mirror hybrid fusion--fission reactor  

SciTech Connect

A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel. (MOW)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1975-06-25T23:59:59.000Z

224

Pure tension superconducting toroidal-field coil system design studies for the Argonne Experimental Power Reactor  

SciTech Connect

As part of the Argonne Tokamak Experimental Power Reactor (TEPR) design studies, a toroidal field (TF) coil system has been designed. NbTi was chosen as the most suitable superconductor and 8T was regarded as a practical peak field level in this study. The 16-coil design was chosen as a reasonable compromise between 2 percent field ripple and 3 m access gap. To minimize the coil structure and the bending moments on the conductor, a pure tension coil shape is necessary. A correct approach for determining the pure tension coil profile in a bumpy TF coil system is given. Verification of the pure tension coil by a three- dimensional stress analysis is presented. For coil quench protection, a series- connected scheme is proposed. (auth)

Wang, S.T.; Purcell, J.R.; Demichele, D.W.; Turner, L.R.

1975-11-01T23:59:59.000Z

225

Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme  

SciTech Connect

In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

Nur Asiah, A.; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2010-06-22T23:59:59.000Z

226

MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS  

SciTech Connect

OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral system scaling analysis, design parameters were obtained and designs of the compact modular 200 MWe SBWR and the full size 1200 MWe SBWR were developed. These reactors are provided with passive safety systems. A new passive vacuum breaker check valve was designed to replace the mechanical vacuum beaker check valve. The new vacuum breaker check valve was based on a hydrostatic head, and was fail safe. The performance of this new valve was evaluated both by the thermal-hydraulic code RELAP5 and by the experiments in a scaled SBWR facility, PUMA. In the core neutronic design a core depletion model was implemented to PARCS code. A lattice design for the SBWR fuel assemblies was performed. Design improvements were made to the neutronics/thermal-hydraulics models of SBWR-200 and SBWR-1200, and design analyses of these reactors were performed. The design base accident analysis and evaluation of all the passive safety systems were completed as scheduled in tasks 4 and 5. Initial conditions for the small break loss of coolant accidents (LOCA) and large break LOCA using REALP5 code were obtained. Small and large break LOCA tests were performed and the data was analyzed. An anticipated transient with scram was simulated using the RELAP5 code for SBWR-200. The transient considered was an accidental closure of the main steam isolation valve (MSIV), which was considered to be the most significant transient. The evaluation of the RELAP5 code against experimental data for SBWR-1200 was completed. In task 6, the instability analysis for the three SBWR designs (SBWR-1200, SBWR-600 and SBWR-200) were simulated for start-up transients and the results were similar. Neither the geysering instability, nor the loop type instability was predicted by RAMONA-4B in the startup simulation following the recommended procedure by GE. The density wave oscillation was not observed at all because the power level used in the simulation was not high enough. A study was made of the potential instabilities by imposing an unrealistically high power ramp in a short time period, as suggested by GE. RAMON

M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

2003-06-16T23:59:59.000Z

227

DESIGN OF A TOKAMAK FUSION REACTOR FIRST WALL ARMOR AGAINST NEUTRAL BEAM IMPINGEMENT  

E-Print Network (OSTI)

Hoffman, et. a1. , "Fusion Reactor First Wall Cooling foricance in Fusion Reactors," Fifth Conference Proceedings onfor a Thp.rmonuclear Reactor," Nu'clear Fusion, 26. H.A.B.

Myers, Richard Allen

2011-01-01T23:59:59.000Z

228

DESIGN OF A TOKAMAK FUSION REACTOR FIRST WALL ARMOR AGAINST NEUTRAL BEAM IMPINGEMENT  

E-Print Network (OSTI)

of Niobium," BNES Nuclear Fusion Reactor Conference, CulhamWall Erosion in Fusion Reactors," Nuclear Fusion. g. 31.and Reactors," Fifth Conference Pro- ceedings on Plasma Physics and Controlled Nuclear Fusion

Myers, Richard Allen

2011-01-01T23:59:59.000Z

229

Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design  

DOE Green Energy (OSTI)

The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal tube film evaporation design used successfully with the BN-350 nuclear plant in Aktau, Kazakhstan. Parametric studies have been performed to optimize the balance of plant design. Also, an economic analysis has been performed, which shows that IRIS-D should be able to provide electricity and clean water at highly competitive costs.

Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

2004-10-06T23:59:59.000Z

230

Newest Los Alamos facility receives LEED® Gold certification  

NLE Websites -- All DOE Office Websites (Extended Search)

is first to achieve both the Leadership in Energy and Environmental Design status and LEED Gold certification. June 13, 2012 Radiological Laboratory Utility Office Building...

231

Thermal Hydraulics of Sodium-Cooled Fast Reactors: Key Design and Safety Issues and Highlights  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Fission Reactors; Thermal Hydraulics

Hisashi Ninokata; Hideki Kamide

232

FEASIBILITY STUDY FOR THE CONCEPTUAL DESIGN OF ADUAL-CORE BOILING SUPERHEAT REACTOR.  

E-Print Network (OSTI)

??For research concerning economical applications of high temperature reactortechnology, a novel approach for creating a Boiling Superheat Reactor (BSR) byaugmenting an Advanced Boiling Water Reactor… (more)

Ross, Jacob

2009-01-01T23:59:59.000Z

233

Fuel cycle design and analysis of SABR: subrcritical advanced burner reactor.  

E-Print Network (OSTI)

??Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been… (more)

Sommer, Christopher

2008-01-01T23:59:59.000Z

234

NERSC Grid Certificates  

NLE Websites -- All DOE Office Websites (Extended Search)

Grid Grid Certificates Grid Certificates | Tags: Grid Grid certificates allow you to access NERSC (and other Grid enabled computing facilities) via grid interfaces. Grid certificates are credentials that must be initialized for use with grid tools. Once a certificate is initialized it is automatically used by the grid tools to authenticate the user to the grid resource. Getting a Short Lived NERSC CA Certificate The NERSC Online CA now offers a quick and painless way to obtain grid certificates. You can obtain a grid certificate with a single command using this method. If you are on a NERSC system, load the globus module to set up your environment: % module load globus or % module load osg On the client system (assuming you have the globus binaries in your path), simply run:

235

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report  

DOE Green Energy (OSTI)

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.

Bruce G. Schnitzler

2012-01-01T23:59:59.000Z

236

Involvement of Certification Agencies  

Science Conference Proceedings (OSTI)

...are also used with other specific meanings, MIL-HDBK- 17 (qualification and certification) or a government

237

ENERGY STAR certification for your building | ENERGY STAR Buildings &  

NLE Websites -- All DOE Office Websites (Extended Search)

certification for your building certification for your building Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In this section Learn the benefits Get started Use Portfolio Manager Save energy Find financing Earn recognition 20-percent recognition ENERGY STAR certification How to apply for ENERGY STAR certification Tips for low-cost verifications Submit a profile of your building

238

ENERGY STAR certification | ENERGY STAR Buildings & Plants  

NLE Websites -- All DOE Office Websites (Extended Search)

ENERGY STAR certification ENERGY STAR certification Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In this section How can we help you? Find out who's partnered with ENERGY STAR Become an ENERGY STAR partner Find ENERGY STAR certified buildings and plants ENERGY STAR certification Featured research and reports Facts and stats Climate change and buildings ENERGY STAR certification

239

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion  

SciTech Connect

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

Bruce G. Schnitzler; Stanley K. Borowski

2012-07-01T23:59:59.000Z

240

Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project  

Science Conference Proceedings (OSTI)

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

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241

Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project  

Science Conference Proceedings (OSTI)

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

242

Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project  

Science Conference Proceedings (OSTI)

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

243

Development of a Scatter Search Optimization Algorithm for Boiling Water Reactor Fuel Lattice Design  

Science Conference Proceedings (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

Juan-Luis François; Cecilia Martín-del-Campo; Luis B. Morales; Miguel-Angel Palomera

244

Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel  

SciTech Connect

The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information Questionnaire (DIQ), prepared by the facility operator, typically with the support of the facility designer. The IAEA will verify design information over the life of the project. This design information is an important IAEA safeguards tool. Since the main interlocutor with the IAEA in each country is the State Regulatory Authority/SSAC (or Regional Regulatory Authority, e.g. EURATOM), the responsibility for conveying this design information to the IAEA falls to the State Regulatory Authority/SSAC.

Philip Casey Durst; Mark Schanfein

2012-08-01T23:59:59.000Z

245

Conversion of methanol to light olefins on SAPO-34: kinetic modeling and reactor design  

E-Print Network (OSTI)

In this work, the reaction scheme of the MTO process was written in terms of elementary steps and generated by means of a computer algorithm characterizing the various species by vectors and Boolean relation matrices. The number of rate parameters is very large. To reduce this number the rate parameters related to the steps on the acid sites of the catalyst were modeled in terms of transition state theory and statistical thermodynamics. Use was made of the single event concept to account for the effect of structure of reactant and activated complex on the frequency factor of the rate coefficient of an elementary step. The Evans-Polanyi relation was also utilized to account for the effect of the structure on the change in enthalpy. The structure was determined by means of quantum chemical software. The number of rate parameters of the complete reaction scheme to be determined from experimental data is thus reduced from 726 to 30. Their values were obtained from the experimental data of Abraha by means of a genetic algorithm involving the Levenberg-Marquardt algorithm and combined with sequential quadratic programming. The retained model yields an excellent fit of the experimental data. All the parameters satisfy the statistical tests as well as the rules of carbenium ion chemistry. The kinetic model also reproduces the experimental data of Marchi and Froment, also obtained on SAPO-34. Another set of their data was used to introduce the deactivation of the catalyst into the kinetic equations. This detailed kinetic model was used to investigate the influence of the operating conditions on the product distribution in a multi-bed adiabatic reactor with plug flow. It was further inserted into riser and fluidized bed reactor models to study the conceptual design of an MTO reactor, accounting for the strong exothermicity of the process. Multi-bed adiabatic and fluidized bed technologies show good potential for the industrial process for the conversion of methanol into olefins.

Al Wahabi, Saeed M. H.

2003-12-01T23:59:59.000Z

246

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01T23:59:59.000Z

247

Design and construction of a 7,500 liter immobilized cell reactor-separator for ethanol production from whey  

DOE Green Energy (OSTI)

A 7,500 liter reactor/separator has been constructed for the production of ethanol from concentrated whey permeate. This unit is sited in Hopkinton IA, across the street from a whey generating cheese plant A two phase construction project consisting of (1) building and testing a reactor/separator with a solvent absorber in a single unified housing, and (2) building and testing an extractive distillation/product stripper for the recovery of anhydrous ethanol is under way. The design capacity of this unit is 250,000 gal/yr of anhydrous product. Design and construction details of the reactor/absorber separator are given, and design parameters for the extractive distillation system are described.

Dale, M.C.

1992-12-31T23:59:59.000Z

248

Modular pebble-bed reactor reforming plant design for process heat  

Science Conference Proceedings (OSTI)

This report describes a preliminary design study of a Modular Pebble-Bed Reactor System Reforming (MPB-R) Plant. The system uses one pressure vessel for the reactor and a second pressure vessel for the components, i.e., reformer, steam generator and coolant circulator. The two vessels are connected by coaxial pipes in an arrangement known as the side-by-side (SBS). The goal of the study is to gain an understanding of this particular system and to identify any technical issues that must be resolved for its application to a modular reformer plant. The basic conditions for the MPB-R were selected in common with those of the current study of the MRS-R in-line prismatic fuel concept, specifically, the module core power of 250 MWt, average core power density of 4.1 w/cc, low enriched uranium (LEU) fuel with a /sup 235/U content of 20% homogeneously mixed with thorium, and a target burnup of 80,000 MWD/MT. Study results include the pebble-bed core neutronics and thermal-hydraulic calculations. Core characteristics for both the once-through-then-out (OTTO) and recirculation of fuel sphere refueling schemes were developed. The plant heat balance was calculated with 55% of core power allotted to the reformer.

Lutz, D.E.; Cowan, C.L.; Davis, C.R.; El Sheikh, K.A.; Hui, M.M.; Lipps, A.J.; Wu, T.

1982-09-01T23:59:59.000Z

249

Conceptual Design of a Reactor Pressure Vessel and its Internals for a HPLWR  

Science Conference Proceedings (OSTI)

A design for the Reactor Pressure Vessel (RPV) and its internals for a HPLWR (High Performance Light Water Reactor) is presented. The RPV has been dimensioned using the pressure vessel code for nuclear power plants in Germany. In order to use conventional vessel materials such as 20 MnMoNi 5 5 (United States: SA 508), the vessel inner wall has to be kept only in contact with coolant at inlet temperature. Therefore, the hot coolant pipe connection from the steam plenum to the outlet is separated from the RPV inner wall using a thermal sleeve. The core inside the vessel rests on a support plate which is connected to the core barrel. The steam plenum is fixed on top of the core using support brackets which are attached to the adjustable steam outlet pipes. This way, the steam plenum rests on the outlet flanges of the lower vessel, while the core barrel is suspended at the closure head flange of the vessel to control thermal expansions between the internals and the RPV and to minimize thermal stresses. Both, inlet and outlet mass flows are separated via C-ring seals to prevent mixing. The control rod guides in the upper plenum are also suspended at the vessel flange and aligned inside the core barrel using centering pins. (authors)

Fischer, Kai [EnBW Kraftwerke AG, Kernkraftwerk Philippsburg, Rheinschanzinsel D-76661 Philippsburg (Germany); Starflinger, Joerg; Schulenberg, Thomas [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies P.O. Box 3640, D-76021 Karlsruhe (Germany)

2006-07-01T23:59:59.000Z

250

Helium turbine design for a 1000 MWe gas-cooled fast breeder reactor with closed gas turbine cycle  

SciTech Connect

This report deals exclusively with the preliminary design of a double-flooded helium turbine for a 1000 MWe gas-cooled fast breeder reactor. The influence is studied of several parameters, such as hub ratio, exit angle of the turbine wheel and inlet angle of the guide wheel, on the designed size of the turbine and the centrifugal stress of the blading, in order to get a survey which is helpful in the preliminary design.

Savatteri, C.

1973-02-15T23:59:59.000Z

251

Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report  

SciTech Connect

Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

1997-04-18T23:59:59.000Z

252

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

SciTech Connect

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

253

SINGLE-FLUID TWO-REGION AQUEOUS HOMOGENEOUS REACTOR POWER PLANT CONCEPTUAL DESIGN AND FEASIBILITY STUDY. Final Report  

SciTech Connect

The feasibility of a 150 Mwe aqueous homogeneous nuclear power plant was investigated by a joint study team of the Nuclear Power Group and The Babcock & Wilcox Company. In this concept, the reactor is a single-fluid two-region design in which the fuel solution circulates through the thoria pellet blanket as the coolant. Componeats and plant arrangement were designed to provide maximum overhead accessibility for maintenance. All components in contact with reactor fuel at high pressure are themselves enclosed in close-fitting high-pressure containment envelopes. (auth)

1957-07-01T23:59:59.000Z

254

Home Energy Professional Certifications (Fact Sheet)  

SciTech Connect

As the weatherization and home energy upgrade industries expand and gain recognition, the need for a qualified workforce becomes more apparent. The certification component of the Guidelines project was designed to create meaningful and lasting careers for weatherization workers. Intended for experienced home energy professionals, the four new certifications focus on the most common jobs in the industry: energy auditor, retrofit installer technician, crew leader, and quality control inspector.

Not Available

2012-11-01T23:59:59.000Z

255

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

256

Thermal Design of an Ultrahigh Temperature Vapor Core Reactor Combined Cycle Nuclear Power Plant  

SciTech Connect

Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas/vapor phase nuclear fuel is uniformly mixed with the topping cycle working fluid. Heat is generated homogeneously throughout the working fluid thus extending the metallurgical temperature limit. Because of the high temperature, magnetohydrodynamic (MHD) generation is employed for topping cycle power extraction. MHD rejected heat is transported via compact heat exchanger to a conventional Brayton gas turbine bottoming cycle. High bottoming cycle mass flow rates are required to remove the waste heat because of low heat capacities for the bottoming cycle gas. High mass flow is also necessary to balance the high Uranium Tetrafluoride (UF{sub 4}) mass flow rate in the topping cycle. Heat exchanger design is critical due to the high temperatures and corrosive influence of fluoride compounds and fission products existing in VCR/MHD exhaust. Working fluid compositions for the topping cycle include variations of Uranium Tetrafluoride, Helium and various electrical conductivity seeds for the MHD. Bottoming cycle working fluid compositions include variations of Helium and Xenon. Some thought has been given to include liquid metal vapor in the bottoming cycle for a Cheng or evaporative cooled design enhancement. The NASA Glenn Lewis Research Center code Chemical Equilibrium with Applications (CEA) is utilized for evaluating chemical species existing in the gas stream. Work being conducted demonstrates the compact heat exchanger design, utilization of the CEA code, and assessment of different topping and bottoming working fluid compositions. (authors)

Bays, Samuel E.; Anghaie, Samim; Smith, Blair; Knight, Travis [Innovative Space Power and Propulsion Institute, University of Florida, 202 Nuclear Science Building, Gainesville, FL 32611 (United States)

2004-07-01T23:59:59.000Z

257

Design and optimization of a high thermal flux research reactor via Kriging-based algorithm  

E-Print Network (OSTI)

In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

258

System Design and Analysis of a 900-MW(thermal) Lead-Cooled Fast Reactor  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fission Reactors

Sang Ji Kim; Yonghee Kim; Sergi Hong; Chung Ho Cho; Jae-Hyuk Eoh; Jong Bum Kim; Myung Hwan Wi; Kwi Seok Ha; Eui Kwang Kim

259

Characteristics of Spent Fuel from Plutonium Disposition Reactors. Vol. 3: A Westinghouse Pressurized-Water Reactor Design  

Science Conference Proceedings (OSTI)

This report discusses the results of a simulation study involving the burnup of mixed-oxide (MOX) fuel in a Westinghouse pressurized-water reactor (PWR). The MOX was composed of uranium and plutonium oxides, where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program and considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic composition, the activities, and the decay heat, together with the gamma and neutron dose rates are discussed for the spent fuel. For the steady-state situation involving this PWR burning MOX fuel, two burn histories are reported. In one case, an assembly is burned in the reactor for two cycles, and in the second case and assembly is burned for three cycles. Furthermore, assemblies containing wet annular burnable absorbers (WABAs) and assemblies that do not contain WABAs are considered in all cases. The two-cycle cases have a burnup of 35 GWd/t, and the three-cycle cases have a burnup of 52.5 GWd/t.

Murphy, B.D.

1997-07-01T23:59:59.000Z

260

Safeguards-by-Design:Guidance for High Temperature Gas Reactors (HTGRs) With Prismatic Fuel  

Science Conference Proceedings (OSTI)

Introduction and Purpose The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on prismatic fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information Questionnaire (DIQ), prepared by the facility operator, typically with the support of the facility designer. The IAEA will verify design information over the life of the project. This design information is an important IAEA safeguards tool. Since the main interlocutor with the IAEA in each country is the State Regulatory Authority/SSAC (or Regional Regulatory Authority, e.g. EURATOM), the responsibility for conveying this design information to the IAEA falls to the State Regulatory Authority/SSAC. For the nuclear industry to reap the benefits of SBD (i.e. avoid cost overruns and construction schedule slippages), nuclear facility designers and operators should work closely with the State Regulatory Authority and IAEA as soon as a decision is taken to build a new nuclear facility. Ideally, this interaction should begin during the conceptual design phase and continue throughout construction and start-up of a nuclear facility. Such early coordination and planning could influence decisions on the design of the nuclear material processing flow-sheet, material storage and handling arrangements, and facility layout (including safeguards equipment), etc.

Mark Schanfein; Casey Durst

2012-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

2009-01-01T23:59:59.000Z

262

Preliminary conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR). Status report, January 1978--March 1978  

DOE Green Energy (OSTI)

The DTHR preliminary conceptual design consists of a magnetically confined fusion reactor fitted with a fertile thorium blanket. The fusion driver concept is based on a beam driven plasma, but at sufficiently high plasma densities that neutrons originating from the interactions of bulk plasma ions contribute significantly to the wall loading. The tokamak has a major radius of 5.2 m, a minor radius of 1.2 m, and the elongation is 1.6. All of the magnetic coil systems are superconducting Nb/sub 3/Sn based on the Large Coil Project (LCP) technology. The toroidal field (TF) coils employ an innovative concept, the ''compact D'' configuration. An engineered bundle divertor concept has been developed based on the bundle divertor design techniques developed for TNS and ISX-B. A thermal power of 150MW of 200 keV deuterium is injected into the plasma through six ducts of a positive ion, neutral beam injection system (NBIS). A water cooled, 316 stainless steel vacuum vessel concept was developed and initial scoping analyses look encouraging. The fusile fuel handling system was evaluated and defined. Details of the tritium injection system remain to be developed. Tritium breeding will be assessed in subsequent phases of the DTHR operation. The fusion driver provides a neutron first wall loading of 2MW/m/sup 2/ for fissile production in the blanket.

Kelly, J.L. (ed.)

1978-03-01T23:59:59.000Z

263

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation  

Science Conference Proceedings (OSTI)

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

Chen, Z. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China); Chen, Y.; Bai, Y.; Wang, W.; Chen, Z.; Hu, L.; Long, P. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China)

2012-07-01T23:59:59.000Z

264

Manufacturing Skills Certification System  

Science Conference Proceedings (OSTI)

... system to their business so that they utilize the skills certification system ... provide input to The Manufacturing Institute about aggregate skill needs of ...

2012-09-20T23:59:59.000Z

265

Acquisition Certifications Program Handbook  

Energy.gov (U.S. Department of Energy (DOE))

The Department of Energy (DOE) Acquisition Career Management Program (ACMP) was established to provide a formal, structured approach to certification of DOE’s acquisition workforce.

266

Patent Certification | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Patent Certification Patent Certification SF-2050.11, Patent Certification More Documents & Publications DOE F 2050.11 C:FormsDOE F 2050.11.cdr DOE F 2050.10...

267

Program on Technology Innovation: Technical Support for GE Economic Simplified Boiling Water Reactor (ESBWR)-Radwaste System Design  

Science Conference Proceedings (OSTI)

EPRI has undertaken a review of advanced nuclear plant (ANP) radioactive waste system designs. This work updates EPRI's Utility Requirements Document (URD) for the design of Advanced Light Water Reactor plants. The goal is to capture the radwaste elements that have led to the dramatic improvement in radioactive waste processing in terms of technology advances and improved operating strategy seen in today's operating U.S. plants. This work will form the basis for radioactive waste processing systems in th...

2006-11-21T23:59:59.000Z

268

Design of a boiling water reactor equilibrium core using thorium-uranium fuel  

SciTech Connect

In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

2004-10-06T23:59:59.000Z

269

Evolution of Design Methodologies for Next Generation of Reactor Pressure Vessels and Extensive Role of Thermal-Hydraulic Numerical Tools  

SciTech Connect

The thermal-hydraulic design of the first pressurized water reactors was mainly based on an experimental approach, with a large series of tests on the main equipment [control rod guide tubes, reactor pressure vessel (RPV) plenums, etc.] to check performance.Development of computational fluid dynamics codes and computers now allows for complex simulations of hydraulics phenomena. Provided adequate qualification, these numerical tools are an efficient means to determine hydraulics in the given design and to perform sensitivities for optimization of new designs. Experiments always play their role, first for qualification and then for validation at the last stage of the design. The design of the European Pressurized Water Reactor (EPR), jointly developed by Framatome ANP, Electricite de France (EDF), and the German utilities, is based on both hydraulics calculations and experiments handled in a complementary approach.This paper describes the collective effort launched by Framatome ANP and EDF on hydraulics calculations for the RPV of the EPR. It concerns three-dimensional calculations of RPV inlets, including the cold legs, the RPV downcomer and lower plenum, and the RPV upper plenum up to and including the hot legs. It covers normal operating conditions but also accidental conditions such as pressurized thermal shock in a small-break loss-of-coolant accident. Those hydraulics studies have provided much useful information for the mechanical design of RPV internals.

Bellet, Serge [Electricite de France - Septen (EDF) (France); Goreaud, Nicolas [Framatome ANP(France); Nicaise, Norbert [Framatome ANP (France)

2005-11-15T23:59:59.000Z

270

Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

S. Blaine Grover

2009-05-01T23:59:59.000Z

271

Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel  

SciTech Connect

Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

Rearden, B.T. [Oak Ridge National Laboratory (United States); Anderson, W.J. [Framatome ANP, Inc. (France); Harms, G.A. [Sandia National Laboratories (United States)

2005-08-15T23:59:59.000Z

272

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input  

SciTech Connect

In this study a feasibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850 deg. C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticality was obtained for this reactor.

Meriyanti; Su'ud, Zaki; Rijal, K. [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Zuhair; Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2010-06-22T23:59:59.000Z

273

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities  

SciTech Connect

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

Michael A. Pope

2011-10-01T23:59:59.000Z

274

Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2009-11-01T23:59:59.000Z

275

Graduate Certificate in Transportation  

E-Print Network (OSTI)

Graduate Certificate in Transportation Nohad A. Toulan School of Urban Studies and Planning of Engineering and Computer Science integrated transportation systems. The Graduate Certificate in Transportation their capabilities. Students in the program can choose among a wide range of relevant courses in transportation

Bertini, Robert L.

276

Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project  

SciTech Connect

This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

Brian G. Williams; Jim C. P. Liou; Hiral Kadakia; Bill Phoenix; Richard R. Schultz

2007-02-27T23:59:59.000Z

277

Thermal-Structural Design of a Water Shield For Surface Reactor Missions  

SciTech Connect

Water shielding is an attractive option for an affordable lunar surface fission reactor program. The attractiveness of the water shielding option arises from the relative ease of proto-typing and ground testing, the relatively low development effort needed, as well as the fabrication and operating experience with stainless steel and water. The most significant limitation in using a water shield is temperature: to prevent the formation of voids and the consequent loss of cooling, the water temperature has to be maintained below the saturation temperature corresponding to the shield pressure. This paper examines natural convection for a prototypic water shield design using the computational fluid dynamics (CFD) code CFX-5 as well as analytical modeling. The results show that natural convection is adequate to keep the water well-mixed. The results also show that for the above-ground configuration, shield surface and water temperatures during lunar day conditions are high enough to require shield pressures up to 2.5 atm to prevent void formation. For the buried configuration, a set of ammonia heat pipes attached to the shield outer wall can be used to maintain water temperatures within acceptable limits. Overall the results show that water shielding is feasible for lunar surface applications. The results of the CFD analyses can also be used to guide development of testing plans for shield thermal testing. (authors)

Sadasivan, Pratap; Kapernick, Richard J.; Poston, David I. [D-5 Nuclear Systems Design Group MS K575, Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States)

2006-07-01T23:59:59.000Z

278

Certification | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Certification Certification Certification THE CERTIFICATION PROCESS You Are Sponsored and Associated With the PMCDP Program. How Do You Proceed? Start by taking an inventory of your past project management training, developmental assignments, and experience. Organize the information by the applicable knowledge, skill, and ability as shown in the Certification and Equivalency Guide (CEG), as well as the training curriculum requirements set forth in the Requirements at a Glance Table, or the PMCDP Curriculum Map. Profile Data Entry Log onto ESS. Click on PMCDP at the top of the screen. A drop down menu will appear. Choose "Instructions" and read through the applicable Tutorial chapters before you begin entering data. Once you are comfortable with the data requirements, begin entering your profile information into

279

Evaluation of tubular reactor designs for supercritical water oxidation of U.S. Department of Energy mixed waste  

Science Conference Proceedings (OSTI)

Supercritical water oxidation (SCWO) is an emerging technology for industrial waste treatment and is being developed for treatment of the US Department of Energy (DOE) mixed hazardous and radioactive wastes. In the SCWO process, wastes containing organic material are oxidized in the presence of water at conditions of temperature and pressure above the critical point of water, 374 C and 22.1 MPa. DOE mixed wastes consist of a broad spectrum of liquids, sludges, and solids containing a wide variety of organic components plus inorganic components including radionuclides. This report is a review and evaluation of tubular reactor designs for supercritical water oxidation of US Department of Energy mixed waste. Tubular reactors are evaluated against requirements for treatment of US Department of Energy mixed waste. Requirements that play major roles in the evaluation include achieving acceptable corrosion, deposition, and heat removal rates. A general evaluation is made of tubular reactors and specific reactors are discussed. Based on the evaluations, recommendations are made regarding continued development of supercritical water oxidation reactors for US Department of Energy mixed waste.

Barnes, C.M.

1994-12-01T23:59:59.000Z

280

Design of GA thermochemical water-splitting process for the Mirror Advanced Reactor System  

DOE Green Energy (OSTI)

GA interfaced the sulfur-iodine thermochemical water-splitting cycle to the Mirror Advanced Reactor System (MARS). The results of this effort follow as one section and part of a second section to be included in the MARS final report. This section describes the process and its interface to the reactor. The capital and operating costs for the hydrogen plant are described.

Brown, L.C.

1983-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Preliminary Design For Conventional and Compact Secondary Heat Exchanger in a Molten Salt Reactor  

Science Conference Proceedings (OSTI)

The strategic goal of the Advance Reactors such as AHTR is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors

Piyush Sabharwall; Mike Patterson; Ali Siahpush; Eung Soo Kim

2012-07-01T23:59:59.000Z

282

Fusion reactor design studies: standard unit costs and cost scaling rules  

SciTech Connect

This report establishes standard unit costs and scaling rules for estimating costs of material, equipment, land, and labor components used in magnetic confinement fusion reactor plant construction and operation. Use of the standard unit costs and scaling rules will add uniformity to cost estimates, and thus allow valid comparison of the economic characteristics of various reactor concepts.

Schulte, S.C.; Bickford, W.E.; Willingham, C.E.; Ghose, S.K.; Walker, M.G.

1979-09-01T23:59:59.000Z

283

Obama Administration Announces $450 Million to Design and Commercialize  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

450 Million to Design and 450 Million to Design and Commercialize U.S. Small Modular Nuclear Reactors Obama Administration Announces $450 Million to Design and Commercialize U.S. Small Modular Nuclear Reactors March 22, 2012 - 2:28pm Addthis COLUMBUS, Ohio - Today, as President Obama went to Ohio State University to discuss the all-out, all-of-the-above strategy for American energy, the White House announced new funding to advance the development of American-made small modular reactors (SMRs), an important element of the President's energy strategy. A total of $450 million will be made available to support first-of-its-kind engineering, design certification and licensing for up to two SMR designs over five years, subject to congressional appropriations. Manufacturing these reactors domestically

284

Obama Administration Announces $450 Million to Design and Commercialize  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

$450 Million to Design and $450 Million to Design and Commercialize U.S. Small Modular Nuclear Reactors Obama Administration Announces $450 Million to Design and Commercialize U.S. Small Modular Nuclear Reactors March 22, 2012 - 2:15pm Addthis Today, as President Obama went to Ohio State University to discuss the all-out, all-of-the-above strategy for American energy, the White House announced new funding to advance the development of American-made small modular reactors (SMRs), an important element of the President's energy strategy. A total of $450 million will be made available to support first-of-its-kind engineering, design certification and licensing for up to two SMR designs over five years, subject to congressional appropriations. Manufacturing these reactors domestically will offer the United States

285

Determination of Optimal Process Flowrates and Reactor Design for Autothermal Hydrogen Production in a Heat-Integrated Ceramic Microchannel Network  

E-Print Network (OSTI)

The present work aimed at designing a thermally efficient microreactor system coupling methanol steam reforming with methanol combustion for autothermal hydrogen production. A preliminary study was performed by analyzing three prototype reactor configurations to identify the optimal radial distribution pattern upon enhancing the reactor self-insulation. The annular heat integration pattern of Architecture C showed superior performance in providing efficient heat retention to the system with a 50 - 150 degrees C decrease in maximum external-surface temperature. Detailed work was performed using Architecture C configuration to optimize the catalyst placement in the microreactor network, and optimize reforming and combustion flows, using no third coolant line. The optimized combustion and reforming catalyst configuration prevented the hot-spot migration from the reactor midpoint and enabled stable reactor operation at all process flowrates studied. Best results were obtained at high reforming flowrates (1800 sccm) with an increase in combustion flowrate (300 sccm) with the net H2 yield of 53% and thermal efficiency of >80% from methanol with minimal insulation to the heatintegrated microchannel network. The use of the third bank of channels for recuperative heat exchange by four different reactor configurations was explored to further enhance the reactor performance; the maximum overall hydrogen yield was increased to 58% by preheating the reforming stream in the outer 16 heat retention channels. An initial 3-D COMSOL model of the 25-channeled heat-exchanger microreactor was developed to predict the reactor hotspot shape, location, optimum process flowrates and substrate thermal conductivity. This study indicated that low thermal conductivity materials (e.g. ceramics, glass) provides enhanced efficiencies than high conductivity materials (e.g. silicon, stainless steel), by maintaining substantial thermal gradients in the system through minimization of axial heat conduction. Final summary of the study included the determination of system energy density; a gravimetric energy density of 169.34 Wh/kg and a volumetric energy density of 506.02 Wh/l were achieved from brass architectures for 10 hrs operation, which is higher than the energy density of Li-Ion batteries (120 Wh/kg and 350 Wh/l). Overall, this research successfully established the optimal process flowrates and reactor design to enhance the potential of a thermally-efficient heat-exchanger microchannel network for autothermal hydrogen production in portable applications.

Damodharan, Shalini

2012-05-01T23:59:59.000Z

286

Design of a continuous-flow reactor for in situ x-ray absorption spectroscopy of solids in supercritical fluids  

Science Conference Proceedings (OSTI)

This paper presents the design and performance of a novel high-temperature and high-pressure continuous-flow reactor, which allows for x-ray absorption spectroscopy or diffraction in supercritical water and other fluids under high pressure and temperature. The in situ cell consists of a tube of sintered, polycrystalline aluminum nitride, which is tolerant to corrosive chemical media, and was designed to be stable at temperatures up to 500 deg. C and pressures up to 30 MPa. The performance of the reactor is demonstrated by the measurement of extended x-ray absorption fine structure spectra of a carbon-supported ruthenium catalyst during the continuous hydrothermal gasification of ethanol in supercritical water at 400 deg. C and 24 MPa.

Dreher, M.; De Boni, E.; Nachtegaal, M.; Wambach, J.; Vogel, F. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland)

2012-05-15T23:59:59.000Z

287

Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012  

SciTech Connect

Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

David W. Nigg; Sean R. Morrell

2012-09-01T23:59:59.000Z

288

Design for Ceramic Membrane Reactor with two Reactant Gases at Different Pressures  

DOE Patents (OSTI)

The invention is a ceramic membrane reactor for syngas production having a reaction chamber, an inlet in the reactor for natural gas intake, a plurality of oxygen permeating ceramic slabs inside the reaction chamber with each slab having a plurality of passages paralleling the gas flow for transporting air through the reaction chamber, a manifold affixed to one end of the reaction chamber for intake of air connected to the slabs, a second manifold affixed to the reactor for removing the oxygen depleted air, and an outlet in the reaction chamber for removing syngas.

Balachandran, Uthamalingam; Mieville, Rodney L.

1998-11-18T23:59:59.000Z

289

PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP  

Science Conference Proceedings (OSTI)

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01T23:59:59.000Z

290

Improved Design of Nuclear Reactor Control System | U.S. DOE...  

Office of Science (SC) Website

energy released during the neutron-induced fission of nuclear fuels is used for energy production in power reactors. The process of beta-n emission from fission products...

291

Optimized core design of a supercritical carbon dioxide-cooled fast reactor  

E-Print Network (OSTI)

Spurred by the renewed interest in nuclear power, Gas-cooled Fast Reactors (GFRs) have received increasing attention in the past decade. Motivated by the goals of the Generation-IV International Forum (GIF), a GFR cooled ...

Handwerk, Christopher S. (Christopher Stanley), 1974-

2007-01-01T23:59:59.000Z

292

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

293

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

294

Design, analysis and optimization of the power conversion system for the Modular Pebble Bed Reactor System  

E-Print Network (OSTI)

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a GenIV nuclear system. The availability of controllable ...

Wang, Chunyun, 1968-

2003-01-01T23:59:59.000Z

295

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

Reactors for Enhanced Nuclear Energy Sustainability”. In:for low-waste proliferation- resistant nuclear energy”.In: Progress in Nuclear Energy 40.3-4 (2002), pp. 431–439. [

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

296

Optimal design and observation of counter-current autothermal reactors for the production of hydrogen  

Science Conference Proceedings (OSTI)

Autothermal reactors, coupling endothermic and exothermic reactions in parallel channels, represent one of the most promising technologies for hydrogen production. Building our prior results, the present work focuses on hydrogen generation in counter-current ...

Michael Baldea; Monica Zanfir; Prodromos Daoutidis

2009-06-01T23:59:59.000Z

297

Thermal hydraulic design and analysis of a large lead-cooled reactor with flexible conversion ratio  

E-Print Network (OSTI)

This thesis contributes to the Flexible Conversion Ratio Fast Reactor Systems Evaluation Project, a part of the Nuclear Cycle Technology and Policy Program funded by the Department of Energy through the Nuclear Energy ...

Nikiforova, Anna S., S.M. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

298

Newest Los Alamos facility receives LEED® Gold certification  

NLE Websites -- All DOE Office Websites (Extended Search)

Newest facility receives LEED® Gold certification Newest facility receives LEED® Gold certification Newest Los Alamos facility receives LEED® Gold certification The Radiological Laboratory Utility Office Building is first to achieve both the Leadership in Energy and Environmental Design status and LEED Gold certification. June 13, 2012 Radiological Laboratory Utility Office Building Radiological Laboratory Utility Office Building Contact Kim Powell Communications Office (505) 695-6159 Email LOS ALAMOS, New Mexico, June 13, 2012-Los Alamos National Laboratory's newest facility, the Radiological Laboratory Utility Office Building (RLUOB), is also its first to achieve both the Leadership in Energy and Environmental Design (LEED) status and LEED Gold certification from the U.S. Green Building Council (USGBC).

299

0707A5 Certificate  

Science Conference Proceedings (OSTI)

Certified Reference Material Address: 2710 S. Boulder Drive, Urbana, IL 61802 USA Email: CRM@aocs.org Tel:+1-217-359-2344; Fax: +1-217-351-8091 ISO Guide 34 A2LA Certificate 3438.01

300

Certification Application Templates  

Energy.gov (U.S. Department of Energy (DOE))

The application templates for each level of FPD certification mirror the competency requirements listed in the CEG. Candidates may use these to prepare their profiles in an editable format prior to...

Note: This page contains sample records for the topic "reactor design certification" 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

Certification plan for safety and PRA codes  

Science Conference Proceedings (OSTI)

A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations.

Toffer, H.; Crowe, R.D. (Westinghouse Hanford Co., Richland, WA (United States)); Ades, M.J. (Westinghouse Savannah River Co., Aiken, SC (United States))

1990-05-01T23:59:59.000Z

302

Certification plan for safety and PRA codes  

Science Conference Proceedings (OSTI)

A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA&PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan`s objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations.

Toffer, H.; Crowe, R.D. [Westinghouse Hanford Co., Richland, WA (United States); Ades, M.J. [Westinghouse Savannah River Co., Aiken, SC (United States)

1990-05-01T23:59:59.000Z

303

SHIELD certification package  

Science Conference Proceedings (OSTI)

Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system built February, 1992, by W.S. Parks. These modules are used for burnup, cooling, separate, and edit calculations.

Boman, C.

1992-02-01T23:59:59.000Z

304

REACTOR FUEL WASTE DISPOSAL PROJECT DEVELOPMENT OF DESIGN PRINCIPLE FOR DISPOSAL OF REACTOR FUEL WASTE INTO UNDERGROUND SALT CAVITIES  

SciTech Connect

Waste disposal in underground salt cavities is considered. Theoretical Investigations for spherical and cylindrical cavities included analysis of elastic stress, thermal stress, and stress redistribution due to the development of a plastic zone around the cavity. The problems of temperature distribution and accompanying thermal stress, due to heat emission from the waste, were also studied. The reduction of the cavity volume, the development of the plastic zone, and the resulting stress redistribution around the cavity are presented as functions of cavity depth, internal pressure of cavity, strenzth of salt, and cavity teraperature rise. It is shown that a salt cavity can be designed such that it is structurally stable as a storage container assuming a chemical equilibrium can be established between the liquid waste and salt. (W.D.M.)

Serata, S.; Gloyna, E.F.

1959-01-01T23:59:59.000Z

305

Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs  

SciTech Connect

This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

Yoder, G.L.

2005-10-03T23:59:59.000Z

306

Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

Blaine Grover

2012-10-01T23:59:59.000Z

307

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

Science Conference Proceedings (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

308

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010  

Science Conference Proceedings (OSTI)

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01T23:59:59.000Z

309

ENERGY STAR Industrial Plant Certification: Instructions for applying |  

NLE Websites -- All DOE Office Websites (Extended Search)

Industrial Plant Certification: Instructions for Industrial Plant Certification: Instructions for applying Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In This Section Campaigns Commercial building design Communications resources Energy management guidance Financial resources Portfolio Manager Products and purchasing Recognition Research and reports Service and product provider (SPP) resources

310

ENERGY STAR Industrial Plant Certification: Professional Engineers' Guide  

NLE Websites -- All DOE Office Websites (Extended Search)

Industrial Plant Certification: Professional Industrial Plant Certification: Professional Engineers' Guide Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In This Section Campaigns Commercial building design Communications resources Energy management guidance Financial resources Portfolio Manager Products and purchasing Recognition Research and reports Service and product provider (SPP) resources

311

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems  

Science Conference Proceedings (OSTI)

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

Ronald G. Ballinger Chunyun Wang Andrew Kadak Neil Todreas

2004-08-30T23:59:59.000Z

312

Reliable-linac design for accelerator-driven subcritical reactor systems.  

Science Conference Proceedings (OSTI)

Accelerator reliability corresponding to a very low frequency of beam interrupts is an important new accelerator requirement for accelerator-driven subcritical reactor systems. In this paper we review typical accelerator-reliability requirements and discuss possible methods for meeting these goals with superconducting proton-linac technology.

Wangler, Thomas P.,

2002-01-01T23:59:59.000Z

313

Investigation of Thermal Feedback Design for Improved Load-Following Capability of Thorium Molten Salt Reactors  

Science Conference Proceedings (OSTI)

The increasing deployment of renewable energy sources has raised concerns about the ramp-rate limitations of conventional steam and combustion turbines in providing load following during solar photovoltaic transients. As one of the promising Generation ... Keywords: molten salt reactors, thorium

Andrew M. Dodson, Roy A. Mccann

2013-04-01T23:59:59.000Z

314

401 Certification (Vermont) | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

401 Certification (Vermont) 401 Certification (Vermont) Eligibility Utility Industrial Savings For Buying & Making Electricity Water Home Weatherization Program Information Vermont...

315

International Comparison of Product Certification and Verification...  

NLE Websites -- All DOE Office Websites (Extended Search)

International Comparison of Product Certification and Verification Methods for Appliances Title International Comparison of Product Certification and Verification Methods for...

316

ACMP Handbook Change 2011-01 - Revised Financial Assistance Certification  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

1-01 - Revised Financial Assistance Certification 1-01 - Revised Financial Assistance Certification 1 Chapter 4 Financial Assistance Certification Financial Assistance Qualification Standards Financial assistance award and administration within DOE is performed by both GS-1102 contracting personnel and non-GS1102 contracting personnel. In order to ensure that personnel in the financial assistance area possess the requisite skill set the Financial Assistance Program (FAP), certification is required. The Financial Assistance Certification Program (FACP) curriculum is designed to provide members with the necessary financial assistance skills and with experience in a wide range of thinking, approaches, and practices. Level IV is for only those individuals involved in the award and administration of Technology

317

ACQUISITION CERTIFICATION - FINANCIAL ASSISTANCE LEVEL I APPLICATION FOR CERTIFICATION  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

ACQUISITION CERTIFICATION - FINANCIAL ASSISTANCE ACQUISITION CERTIFICATION - FINANCIAL ASSISTANCE LEVEL I APPLICATION FOR CERTIFICATION TECHNICAL PROJECT OFFICER LEVEL I PART A - EMPLOYEE INFORMATION Name (Last, First, Middle initial)_____________________________________________ Email Address____________________________________________________________ Phone __________________________________________________________________ Agency Name____________________________________________________________ Agency Address__________________________________________________________ Title, Series, Grade________________________________________________________ Education: Please specify degree and major: a. Degree: Associates: __; Bachelors __; Masters: __; Doctorate: __

318

Certification process of safety analysis and risk management computer codes at the Savannah River Site  

Science Conference Proceedings (OSTI)

The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.

Ades, M.J. (Westinghouse Savannah River Co., Aiken, SC (United States)); Toffer, H.; Lewis, C.J.; Crowe, R.D. (Westinghouse Hanford Co., Richland, WA (United States))

1992-01-01T23:59:59.000Z

319

Certification process of safety analysis and risk management computer codes at the Savannah River Site  

Science Conference Proceedings (OSTI)

The commitment by Westinghouse Savannah River Company (WSRC) to bring safety analysis and risk management codes into compliance with national and sitewide quality assurance requirements necessitated a systematic, structured approach. As a part of this effort, WSRC, in cooperation with the Westinghouse Hanford Company, has developed and implemented a certification process for the development and control of computer software. Safety analysis and risk management computer codes pertinent to reactor analyses were selected for inclusion in the certification process. As a first step, documented plans were developed for implementing verification and validation of the codes, and establishing configuration control. User qualification guidelines were determined. The plans were followed with an extensive assessment of the codes with respect to certification status. Detailed schedules and work plans were thus determined for completing certification of the codes considered. Although the software certification process discussed is specific to the application described, it is sufficiently general to provide useful insights and guidance for certification of other software.

Ades, M.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Toffer, H.; Lewis, C.J.; Crowe, R.D. [Westinghouse Hanford Co., Richland, WA (United States)

1992-05-01T23:59:59.000Z

320

Who Owns Renewable Energy Certificates?  

E-Print Network (OSTI)

Who Owns Renewable Energy Certificates? Edward Holt, RyanME 04079 edholt@igc.org Renewable energy certificates (RECs)convey the attributes of a renewable energy generator and

Holt, Edward; Wiser, Ryan; Bolinger, Mark

2006-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Conceptual design of thorium-fuelled Mitrailleuse accelerator-driven subcritical reactor using D-Be neutron source  

Science Conference Proceedings (OSTI)

A distributed accelerator is a charged-particle accelerator that uses a new acceleration method based on repeated electrostatic acceleration. This method offers outstanding benefits not possible with the conventional radio-frequency acceleration method, including: (1) high acceleration efficiency, (2) large acceleration current, and (3) lower failure rate made possible by a fully solid-state acceleration field generation circuit. A 'Mitrailleuse Accelerator' is a product we have conceived to optimize this distributed accelerator technology for use with a high-strength neutron source. We have completed the conceptual design of a Mitrailleuse Accelerator and of a thorium-fuelled subcritical reactor driven by a Mitrailleuse Accelerator. This paper presents the conceptual design details and approach to implementing the subcritical reactor core. We will spend the next year or so on detailed design work, and then will start work on developing a prototype for demonstration. If there are no obstacles in setting up a development organization, we expect to finish verifying the prototype's performance by the third quarter of 2015. (authors)

Kokubo, Y. [Quan Japan Company Limited, 3-9-15 Sannomiya-cho, Chuo-ku, Kobe, Hyogo, 650-0021 (Japan); Kamei, T. [Research Inst. for Applied Sciences, 49 Tanaka Ohicho, Sakyo-ku, Kyoto-shi, Kyoto, 606-8202 (Japan)

2012-07-01T23:59:59.000Z

322

certification_poststandardchange.pdf  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Certification Following a Change Certification Following a Change in the Applicable Energy or Water Conservation Standard The Energy Policy and Conservation Act of 1975, as amended, (EPCA or the ''Act'') authorizes the U. S. Department of Energy (DOE or the ''Department'') to enforce compliance with the energy conservation standards established for certain consumer products and commercial equipment. 42 U.S.C. §§ 6299-6305 (consumer products), 6316 (commercial and industrial equipment). To ensure that all covered products distributed in the United States comply with DOE's energy conservation standards, the Department has promulgated enforcement regulations, which include specific certification and compliance requirements. See Title 10 of the Code of Federal Regulations (10 CFR) part 430, subpart F; 10 CFR part 431,

323

Conceptual design analysis of an MHD power conversion system for droplet-vapor core reactors. Final report  

DOE Green Energy (OSTI)

A nuclear driven magnetohydrodynamic (MHD) generator system is proposed for the space nuclear applications of few hundreds of megawatts. The MHD generator is coupled to a vapor-droplet core reactor that delivers partially ionized fissioning plasma at temperatures in range of 3,000 to 4,000 K. A detailed MHD model is developed to analyze the basic electrodynamics phenomena and to perform the design analysis of the nuclear driven MHD generator. An incompressible quasi one dimensional model is also developed to perform parametric analyses.

Anghaie, S.; Saraph, G.

1995-12-31T23:59:59.000Z

324

Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design  

Science Conference Proceedings (OSTI)

The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

Smith, C

2010-02-22T23:59:59.000Z

325

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor  

DOE Patents (OSTI)

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

1999-03-23T23:59:59.000Z

326

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor  

DOE Patents (OSTI)

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

1997-12-01T23:59:59.000Z

327

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor  

DOE Patents (OSTI)

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

Crawford, Douglas C. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Hayes, Steven L. (Idaho Falls, ID); Hill, Robert N. (Bolingbrook, IL)

1999-01-01T23:59:59.000Z

328

Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors  

DOE Green Energy (OSTI)

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermal-hydraulic and efficiency points of view.

Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

2005-11-01T23:59:59.000Z

329

Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant  

DOE Green Energy (OSTI)

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood.

Chang H. Oh; Eung Soo Kim; Steven Sherman

2008-04-01T23:59:59.000Z

330

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

331

Compiled reports on the applicability of selected codes and standards to advanced reactors  

Science Conference Proceedings (OSTI)

The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

1994-08-01T23:59:59.000Z

332

Advanced Reactor Development and Technology - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...

333

Non-ferrous Metals Industry Energy Management System Certification...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Non-ferrous Metals Industry Energy Management System Certification Details about China Quality Certification Center and Energy Management System certifications....

334

Design and demonstration of an immobilized-cell fluidized-bed reactor for the efficient production of ethanol  

DOE Green Energy (OSTI)

Initial studies have been carried out using a 4 inch ID fluidized bed reactor (FBR). This medium scale FBR was designed for scale-up. Present performance was compared with results from experiments using smaller FBRs. On-line and off-line measurement systems are also described. Zymomonas mobilis was immobilized in {kappa}-carrageenan at cell loadings of 15--50 g (dry weight) L{sup {minus}1}. The system is designed for determining optimal operation with high conversion and productivity for a variety of conditions including feedstocks, temperature, flow rate, and column sizes (from 2 to 5 meters tall). The demonstration used non-sterile feedstocks containing either industrial (light steep water) or synthetic nutrients and dextrose.

Webb, O.F.; Scott, T.C.; Davison, B.H.; Scott, C.D.

1994-06-01T23:59:59.000Z

335

Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor  

Science Conference Proceedings (OSTI)

The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.

Piyush Sabharwall; Ali Siahpush; Michael McKellar; Michael Patterson; Eung Soo Kim

2012-06-01T23:59:59.000Z

336

ENERGY STAR plant certification | ENERGY STAR Buildings & Plants  

NLE Websites -- All DOE Office Websites (Extended Search)

» ENERGY STAR plant certification » ENERGY STAR plant certification Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In this section Get started with ENERGY STAR Make the business case Build an energy management program Measure, track, and benchmark Improve energy performance Industrial service and product providers Earn recognition ENERGY STAR Partner of the Year Award

337

Toroidal fusion reactor design based on the reversed-field pinch  

DOE Green Energy (OSTI)

The toroidal reversed-field pinch (RFP) achieves gross equilibrium and stability with a combination of high shear and wall stabilization, rather than the imposition of tokamak-like q-constraints. Consequently, confinement is provided primarily by poloidal magnetic fields, poloidal betas as large as approximately 0.58 are obtainable, the high ohmic-heating (toroidal) current densities promise a sole means of heating a D-T plasma to ignition, and the plasma aspect ratio is not limited by stability/equilibrium constraints. A reactor-like plasma model has been developed in order to quantify and to assess the general features of a power system based upon RFP confinement. An ''operating point'' has been generated on the basis of this plasma model and a relatively detailed engineering energy balance. These results are used to generate a conceptual engineering model of the reversed-field pinch reactor (RFPR) which includes a general description of a 750 MWe power plant and the preliminary consideration of vacuum/fueling, first wall, blanket, magnet coils, iron core, and the energy storage/transfer system.

Hagenson, R.L.

1978-07-01T23:59:59.000Z

338

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

339

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

in Nuclear Systems. Tech. rep. Argonne National Laboratory,experiment. Tech. rep. Argonne National Laboratory, 1958. [and Core Design. Tech. rep. Argonne National Labo- ratory.

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

340

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor  

SciTech Connect

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclear pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.

Kim, Seokho H [ORNL; Berry, Jan [ORNL

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Building Operator Certification: Improving Commercial Building Energy Efficiency Through Operator Training and Certification  

E-Print Network (OSTI)

Building Operator Certification (BOC) is a competency-based certification for building operators designed to improve the energy efficiency of commercial buildings. Operators earn certification by attending training sessions and completing project assignments in their facilities. The certification provides a credential for their professional development while also offering employers a way to identify skilled operators. Developed as a market transformation venture with funding from the Northwest Energy Efficiency Alliance, evaluation research has shown BOC participants are saving money and energy in their facilities, and awareness among major employers is growing. BOC is now expanding to serve other regions of the country including the Northeast, California and Wisconsin. It is being offered as a turnkey program to interested organizations and utilities and can be operationalized within a year to serve customers quickly. It also serves as a platform for other energy efficiency initiatives such as building commissioning, EnergyStar benchmarking, and resource conservation manager. This paper will discuss the development and implementation of the Building Operator Certification, market response to BOC in the Northwest and Northeast, and energy saving and customer service benefits.

Putnam, C.; Mulak, A.

2001-01-01T23:59:59.000Z

342

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network (OSTI)

A. L. London, Compact Heat Exchangers 3 rd Edition McGraw-A. L. London, Compact Heat Exchangers 3 rd Edition McGraw-that short and compact NDHX heat exchanger designs can be

Galvez, Cristhian

2011-01-01T23:59:59.000Z

343

Design of compact intermediate heat exchangers for gas cooled fast reactors  

E-Print Network (OSTI)

Two aspects of an intermediate heat exchanger (IHX) for GFR service have been investigated: (1) the intrinsic characteristics of the proposed compact printed circuit heat exchanger (PCHE); and (2) a specific design optimizing ...

Gezelius, Knut, 1978-

2004-01-01T23:59:59.000Z

344

CERTIFICATE OF AUTHENTICITY  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CERTIFICATE OF AUTHENTICITY CERTIFICATE OF AUTHENTICITY I hereby certify that this transcript constitutes an accurate record of the full Council meeting of the National Coal Council held on November 14,2008 at the Westin Grand Hotel, Washington, D.C. ~ i c h a e l G. Mueller, Chair National Coal Council I NATIONAL COAL COUNCIL 1 FULL COUNCIL MEETING FRIDAY NOVEMBER 14, 2 0 0 8 The Full Council meeting convened at 9 : 0 0 a.m. in the Washington Ballroom of the Westin Grand Hotel, 2 3 5 0 M Street, NW, Washington, DC, Chair Michael G. Mueller presiding. ATTENDEES: MIKE MUELLER, Arneren Energy Fuels & Services Company, Chair RICH EIMER, Dynegy Inc., Vice-Chair ROBERT BECK, National Coal Council, Executive Vice-President SY ALI, Clean Energy Consulting BARB ALTIZER, Eastern Coal Council

345

certification_samplingplan.pdf  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Applicable Sampling Plan Applicable Sampling Plan for Certification of Consumer Products The Energy Policy and Conservation Act of 1975, as amended, (EPCA or the ''Act'') authorizes the Department of Energy (DOE or the ''Department'') to enforce compliance with the energy conservation standards established for certain consumer products and commercial equipment. 42 U.S.C. §§ 6299-6305. To ensure that all covered products distributed in the United States comply with DOE's energy conservation standards, the Department has promulgated enforcement regulations, which include specific certification and compliance requirements. See 10 CFR part 430, subpart F. Each manufacturer must certify that each basic model meets the applicable energy conservation standard before distributing that model in commerce. 10 CFR 430.62(a). To

346

TORT certification package  

Science Conference Proceedings (OSTI)

The TORT code has been certified. TORT is a three-dimensional discrete ordinates transport theory code, than can solve neutron, photon, or coupled neutron/photon problems. The code will be used primarily for shielding and radiation field calculations SRS. As defined in this work, certification dies not imply validation. The code must be validated for a particular type of calculation before it can be used for critical applications.

Frost, R.L.

1993-10-01T23:59:59.000Z

347

Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes  

E-Print Network (OSTI)

Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet path is shifted from the annular path between reactor core barrel and vessel wall through the permanent side reflector (PSR). The number and dimensions of coolant holes are varied to optimize the pressure drop, the inlet velocity, and the percentage of graphite removed from the PSR to create this inlet path. With the removal of ~10% of the graphite from PSR the PVT is reduced from 541 0C to 421 0C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant temperature with the aim of further reduction of PVT. The dimensions and number of fuel rods and coolant holes, and the triangular pitch have been changed and optimized. Different packing fractions for the new core design have been used to conserve the number of fuel particles. Thermal properties for the fuel elements are calculated and incorporated into these analyses. The inlet temperature, mass flow and bypass flow are optimized to limit the peak fuel temperature (PFT) within an acceptable range. Using both of these modifications together, the PVT is reduced to ~350 0C while keeping the outlet temperature at 950 0C and maintaining the PFT within acceptable limits. The vessel and fuel temperatures during low pressure conduction cooldown and high pressure conduction cooldown transients are found to be well below the design limits. The reliability and availability studies for coupled nuclear hydrogen production processes based on the sulfur iodine thermochemical process and high temperature electrolysis process have been accomplished. The fault tree models for both these processes are developed. Using information obtained on system configuration, component failure probability, component repair time and system operating modes and conditions, the system reliability and availability are assessed. Required redundancies are made to improve system reliability and to optimize the plant design for economic performance. The failure rates and outage factors of both processes are found to be well below the maximum acceptable range.

Reza, S.M. Mohsin

2007-05-01T23:59:59.000Z

348

Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features  

Science Conference Proceedings (OSTI)

A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

Winters, J. W.; Corletti, M. M.; Hayashi, Y.

2003-11-12T23:59:59.000Z

349

RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor  

SciTech Connect

U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment with no field experience and lowers specific costs while providing higher reliability. This paper presents a brief description of the TCWS conceptual design and the application of RAMI tools to optimize the design at different stages during the project.

Ferrada, Juan J [ORNL; Reiersen, Wayne T [ORNL

2011-01-01T23:59:59.000Z

350

NEUTRONIC REACTORS  

DOE Patents (OSTI)

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Anderson, H.L.

1958-10-01T23:59:59.000Z

351

Monte Carlo methods of neutron beam design for neutron capture therapy at the MIT Research Reactor (MITR-II)  

Science Conference Proceedings (OSTI)

Monte Carlo methods of coupled neutron/photon transport are being used in the design of filtered beams for Neutron Capture Therapy (NCT). This method of beam analysis provides segregation of each individual dose component, and thereby facilitates beam optimization. The Monte Carlo method is discussed in some detail in relation to NCT epithermal beam design. Ideal neutron beams (i.e., plane-wave monoenergetic neutron beams with no primary gamma-ray contamination) have been modeled both for comparison and to establish target conditions for a practical NCT epithermal beam design. Detailed models of the 5 MWt Massachusetts Institute of Technology Research Reactor (MITR-II) together with a polyethylene head phantom have been used to characterize approximately 100 beam filter and moderator configurations. Using the Monte Carlo methodology of beam design and benchmarking/calibrating our computations with measurements, has resulted in an epithermal beam design which is useful for therapy of deep-seated brain tumors. This beam is predicted to be capable of delivering a dose of 2000 RBE-cGy (cJ/kg) to a therapeutic advantage depth of 5.7 cm in polyethylene assuming 30 micrograms/g 10B in tumor with a ten-to-one tumor-to-blood ratio, and a beam diameter of 18.4 cm. The advantage ratio (AR) is predicted to be 2.2 with a total irradiation time of approximately 80 minutes. Further optimization work on the MITR-II epithermal beams is expected to improve the available beams. 20 references.

Clement, S.D.; Choi, J.R.; Zamenhof, R.G.; Yanch, J.C.; Harling, O.K. (Massachusetts Institute of Technology, Cambridge (USA))

1990-01-01T23:59:59.000Z

352

Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

Douglas M. Gerstner

2009-05-01T23:59:59.000Z

353

Equipment Certification | Open Energy Information  

Open Energy Info (EERE)

Equipment Certification Equipment Certification Jump to: navigation, search Policies requiring renewable energy equipment to meet certain standards serve to protect consumers from buying inferior equipment. These requirements not only benefit consumers; they also protect the renewable energy industry by making it more difficult for substandard systems to reach the market. [1] Equipment Certification Incentives CSV (rows 1 - 19) Incentive Incentive Type Place Applicable Sector Eligible Technologies Active Canada Oil and Gas Operations Act (Canada) Environmental Regulations Equipment Certification Fees Generating Facility Rate-Making Generation Disclosure Industry Recruitment/Support Safety and Operational Guidelines Siting and Permitting Canada Commercial Construction Developer

354

Smart Grid Testing and Certification  

Science Conference Proceedings (OSTI)

... ensure the benefits of rigorous test programs are ... Smart Grid products are tested and certified. ... SGIP), particularly the SGIP Testing and Certification ...

2013-04-29T23:59:59.000Z

355

Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Increased Enrichment/High Burnup and Light Water Reactor Fuel Cycle Optimization

B. T. Rearden; W. J. Anderson; G. A. Harms

356

Civil Penalty Actions For Certification Violations Resolved | Department of  

NLE Websites -- All DOE Office Websites (Extended Search)

Civil Penalty Actions For Certification Violations Resolved Civil Penalty Actions For Certification Violations Resolved Civil Penalty Actions For Certification Violations Resolved May 6, 2010 - 12:39pm Addthis Today, the Department of Energy announced that it had resolved civil penalty actions against four showerhead manufacturers for having violated the Department's water conservation certification regulations. To resolve these actions Zoe Industries, Altmans Products LLC, EZ-FLO International, and Watermark Designs Ltd. have now certified that their showerheads meet federal water conservation standards and collectively will pay civil penalties of $165,104. DOE had initiated cases against the four manufacturers on January 25th, alleging that they had failed to submit required documents to DOE demonstrating compliance with federal

357

Civil Penalty Actions For Certification Violations Resolved | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Civil Penalty Actions For Certification Violations Resolved Civil Penalty Actions For Certification Violations Resolved Civil Penalty Actions For Certification Violations Resolved May 6, 2010 - 12:39pm Addthis Today, the Department of Energy announced that it had resolved civil penalty actions against four showerhead manufacturers for having violated the Department's water conservation certification regulations. To resolve these actions Zoe Industries, Altmans Products LLC, EZ-FLO International, and Watermark Designs Ltd. have now certified that their showerheads meet federal water conservation standards and collectively will pay civil penalties of $165,104. DOE had initiated cases against the four manufacturers on January 25th, alleging that they had failed to submit required documents to DOE demonstrating compliance with federal

358

Seismic design technology for breeder reactor structures. Volume 1. Special topics in earthquake ground motion  

Science Conference Proceedings (OSTI)

This report is divided into twelve chapters: seismic hazard analysis procedures, statistical and probabilistic considerations, vertical ground motion characteristics, vertical ground response spectrum shapes, effects of inclined rock strata on site response, correlation of ground response spectra with intensity, intensity attenuation relationships, peak ground acceleration in the very mean field, statistical analysis of response spectral amplitudes, contributions of body and surface waves, evaluation of ground motion characteristics, and design earthquake motions. (DLC)

Reddy, D.P.

1983-04-01T23:59:59.000Z

359

Green Power Network: Renewable Energy Certificates (RECs)  

NLE Websites -- All DOE Office Websites (Extended Search)

Table of Retail Products Table of Retail Products Table of Commercial Certificate Marketers List of REC Marketers REC Prices National Renewable Energy Certificate Tracking Systems Map Carbon Offsets State Policies Renewable Energy Certificates (RECs) Renewable energy certificates (RECs), also known as renewable energy credits, green certificates, green tags, or tradable renewable certificates, represent the environmental attributes of the power produced from renewable energy projects and are sold separate from commodity electricity. Customers can buy green certificates whether or not they have access to green power through their local utility or a competitive electricity marketer. And they can purchase green certificates without having to switch electricity suppliers. Table of Retail Products

360

Building Technologies Office: Implementation, Certification and Enforcement  

NLE Websites -- All DOE Office Websites (Extended Search)

Implementation, Implementation, Certification and Enforcement to someone by E-mail Share Building Technologies Office: Implementation, Certification and Enforcement on Facebook Tweet about Building Technologies Office: Implementation, Certification and Enforcement on Twitter Bookmark Building Technologies Office: Implementation, Certification and Enforcement on Google Bookmark Building Technologies Office: Implementation, Certification and Enforcement on Delicious Rank Building Technologies Office: Implementation, Certification and Enforcement on Digg Find More places to share Building Technologies Office: Implementation, Certification and Enforcement on AddThis.com... About Standards & Test Procedures Implementation, Certification & Enforcement Standardized Templates for Reporting Test Results

Note: This page contains sample records for the topic "reactor design certification" 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

Puerto Rico - Renewable Energy Equipment Certification | Department...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Puerto Rico - Renewable Energy Equipment Certification Puerto Rico - Renewable Energy Equipment Certification Eligibility Construction InstallerContractor Savings For Solar Buying...

362

Certification of solar products - The Florida experience  

SciTech Connect

Florida legislation enacted in 1976 directed the Florida Solar Energy Center (FSEC) to develop standards for solar energy systems manufactured or sold in the state, establish criteria for testing the performance of solar energy systems, and provide a means to display compliance with approved performance tests for these systems. This mandate has been effectively implemented for both solar domestic water heating and solar pool heating systems. With growing interest and markets for photovoltaic systems, plans are presently being developed to expand the scope of the mandate to include photovoltaic technology. This paper discusses four complementary facets of a photovoltaic (PV) system certification program. They include PV module performance characterization and rating; PV system design review and approval; examination and authorization of photovoltaic system installers; and inspection and acceptance testing of PV system installation. The suggested photovoltaic system process builds on lessons learned from over 20 years of testing, certifying and labeling of solar thermal collectors, and the certification of solar thermal systems.

POST,HAROLD N.; ROLAND,JAMES D.; VENTRE,GERARD G.; HUGGINS,JAMES C.

2000-02-02T23:59:59.000Z

363

Primary heat transfer loop design for the Cascade inertial confinement fusion reactor  

Science Conference Proceedings (OSTI)

This study investigates a heat exchanger and balance of plant design to accompany the Cascade inertial confinement fusion reaction chamber concept. The concept uses solid Li/sub 2/O or other lithium-ceramic granules, held to the wall of a rotating reaction chamber by centrifugal action, as a tritium breeding blanket and first wall protection. The Li/sub 2/O granules enter the chamber at 800 K and exit at 1200 K after absorbing the thermal energy produced by the fusion process.

Murray, K.A.; McDowell, M.W.

1984-05-01T23:59:59.000Z

364

DORT certification package  

SciTech Connect

The DORT code has been certified. DORT is a two-dimensional discrete ordinates transport theory code, that can solve neutron, photon, or coupled neutron/photon problems. It is anticipated that DORT will be used for criticality calculations as well as for shielding and radiation field analysis at SRS. In addition to the DORT module itself, 5 utility programs that are useful in certain DORT applications have been certified. These modules are: GIP, DOS, GRTUNCL, BNDRYS, and RTFLUM. As defined in this work, certification does not imply validation. These codes must be validated for a particular type of calculation before they can be used for critical applications.

Frost, R.L.

1994-01-01T23:59:59.000Z

365

Performance demonstration of a high-power space-reactor heat-pipe design  

SciTech Connect

Performance of a 15.9-mm diam, 2-m long, artery heat pipe has been demonstrated at power levels to 22.6 kW and temperatures to 1500/sup 0/K. The heat pipe employed lithium as a working fluid with distribution wicks and arteries fabricated from 400 mesh Mo-41 wt % Re screen. Molybdenum alloy (TZM) was used for the container. Peak axial power density attained in the testing was 19 kW/cm/sup 2/ at 1465/sup 0/K. The corresponding radial flux density in the evaporator region of the heat pipe was 150 W/cm/sup 2/. The extrapolated limit for the heat pipe at its 1500/sup 0/K design point is 30 kW, corresponding to an axial flux density of 25 kW/cm/sup 2/. Sonic and capillary limits for the design were investigated in the 1100 to 1500/sup 0/K temperature range. Excellent agreement of measured and predicted temperature and power levels was observed.

Merrigan, M.A.; Martinez, E.H.; Keddy, E.S.; Runyan, J.; Kemme, J.E.

1983-01-01T23:59:59.000Z

366

Improving the usability of VeriSign® SSL certificate enrollment  

Science Conference Proceedings (OSTI)

This paper discusses the methodology used to redesign the SSL Certificate enrollment process for VeriSign, Inc.Like many technology companies in which time to market is of critical importance, VeriSign's release schedules and fast response to customer ... Keywords: digital security, input and interaction technologies, prototyping, service design, usability testing and evaluation, user interface design

Krista Van Laan; Eva Jonshult

2005-11-01T23:59:59.000Z

367

State-of-the-art review and report on critical aspects and scale-up considerations in the design of fluidized-bed reactors. Final report on Phase 1  

SciTech Connect

Information is given on the design of distributor plates and opening geometry to provide uniform flow over the reactor area. The design of granular bed filters is also considered. Pressure drops and particle size in the bed are discussed. (LTN)

1980-01-01T23:59:59.000Z

368

MCNP certification package  

Science Conference Proceedings (OSTI)

In response to a Department of Energy (DOE) request, Westinghouse Savannah River Company committed to certify all computer codes used in critical calculations at the site. Since the Monte Carlo Neutron Photon Transport (MCNP) code will be used to perform critical analyses involving criticality and shielding, the code must be certified. Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, and establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the MCNP code, version 4.2, built June 11, 1992, by J.H. Hightower on the SRS CRAY. This report does not release MCNP for use under production status for any application for which a MCNP validation document does not exist. These validation documents will describe the specific range of applicability, limitations on use, results and biases for a particular MCNP application.

Trumble, E.F.

1992-08-01T23:59:59.000Z

369

EFFECTS OF THE MAXIMUM CREDIBLE ACCIDENT RELEVANT TO THE DESIGN OF THE CONTAINMENT SHELL, EXPERIMENTAL LOW-TEMPERATURE PROCESS HEAT REACTOR PROJECT  

SciTech Connect

The effects of the maximum credible accident relative to the design of the containment shell are discussed. The maximum credible accident is defined. The thermal and hydraulic effects of the maximum credible accident on the reactor system were analyzed. The extent to which fuelrod cladding will melt was estimated. The amount of energy released from the reactor system by the escaping steam and water and by a possible chemical reaction was calculated along with the corresponding pressure rise inside the containment shell. The kinds, amounts, and total radioactivity of fission products released to the atmosphere of the containment shell after the core melts were predicted. (M.C.G.)

1960-03-21T23:59:59.000Z

370

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors  

E-Print Network (OSTI)

separate effects test steam generators small modular reactorNuclear Generating Station (SONGS) steam generators (SG).January of 2012, a steam generator tube leak was detected,

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

371

Economic characteristics of a smaller, simpler reactor  

SciTech Connect

Reduced load growth and heightened concern with economic risk has led to an expressed utility preference for smaller capacity additions. The Modular High Temperature Reactor (MHTGR) plant has been developed as a small, simple plant that has limited financial risk and is economically competitive with comparatively sized coal plants. Competitive economics is achieved by the simplifications made possible in a small MHTGR, reduction in the quantity of nuclear grade construction and design standardization and certification. Assessments show the MHTGR plant to have an economic advantage over coal plants for plant sizes from 270 MWe to 1080 MWe. Financial risk is limited by small unit sizes and short lead times that allow incremental deployment. Evaluations show the MHTGR incremental deployment capability to reduce negative cash flows by almost a factor of 2 relative to that required by a single large nuclear plant.

LaBar, M.; Bowers, H.

1988-01-01T23:59:59.000Z

372

Directory of Federal Government Certification Programs  

Science Conference Proceedings (OSTI)

Page 1. NBS Special Publication 739 Directory of Federal Government Certification Programs Maureen Breitenberg, Editor ...

2011-07-29T23:59:59.000Z

373

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

Science Conference Proceedings (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

374

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

Science Conference Proceedings (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

375

Solar Equipment Certification | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Solar Equipment Certification Solar Equipment Certification Solar Equipment Certification < Back Eligibility Commercial Construction Installer/Contractor Residential Savings Category Heating & Cooling Commercial Heating & Cooling Solar Heating Water Heating Program Info State Minnesota Program Type Equipment Certification Provider Minnesota Department of Commerce Minnesota law requires that all active solar space-heating and water-heating systems, sold, offered for sale, or installed on residential and commercial buildings in the state meet Solar Rating and Certification Corporation (SRCC) standards. Specifically, the rule references SRCC's "Operating Guidelines" pertaining to collector certification and system certification: OG-100 and OG-300, respectively. Local building officials

376

EPA Redesigns Conversion Certification Policies  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

EPA Redesigns EPA Redesigns Conversion Certification Policies At a recent meeting held in Washington, DC, officials from the U.S. Environmental Protection Agency (EPA) opened dialogue about proposed changes to its emission certification policies that affect alternative fuel vehicles (AFVs). "We are trying to accommo- date the Energy Policy Act (EPAct) and Executive Order requirements while trying to change enforce- ment policies and guidance with respect to conversions," said Rich Ackerman of EPA's Enforcement Office. The meeting, attended by representatives of more than 60 organizations, was held to discuss actions addressing AFV emission certification. Specifically, topics included * Conversion emissions perfor- mance data * Status of environmental laws pertaining to alternative fuel

377

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

378

Renewable Energy Certificate Program  

Science Conference Proceedings (OSTI)

This project was primarily to develop and implement a curriculum which will train undergraduate and graduate students at the University seeking a degree as well as training for enrollees in a special certification program to prepare individuals to be employed in a broad range of occupations in the field of renewable energy and energy conservation. Curriculum development was by teams of Saint Francis University Faculty in the Business Administration and Science Departments and industry experts. Students seeking undergraduate and graduate degrees are able to enroll in courses offered within these departments which will combine theory and hands-on training in the various elements of wind power development. For example, the business department curriculum areas include economic modeling, finance, contracting, etc. The science areas include meteorology, energy conversion and projection, species identification, habitat protection, field data collection and analysis, etc.

Gwendolyn S. Andersen

2012-07-17T23:59:59.000Z

379

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

380

First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

First Step to Spur U.S. Manufacturing of Small Modular Nuclear First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors January 25, 2012 - 5:06pm Addthis Brenda DeGraffenreid The Energy Department recently announced the first step toward manufacturing small modular nuclear reactors (SMRs) in the United States, demonstrating the Administration's commitment to advancing U.S. manufacturing leadership in low-carbon, next generation energy technologies and restarting the nation's nuclear industry. The release of a draft Funding Opportunity Announcement (FOA) last week presents supply-chain procurement opportunities for our nation's small businesses down the line, as industry provides input in advance of a full FOA on engineering, design certification, and licensing through a

Note: This page contains sample records for the topic "reactor design certification" 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

Policy Flash 2013-54 Acquisition Certification Program Handbook...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Policy Flash 2013-54 Acquisition Certification Program Handbook 4th Edition Policy Flash 2013-54 Acquisition Certification Program Handbook 4th Edition Acquisition Certifications...

382

LEED Certification | Open Energy Information  

Open Energy Info (EERE)

Certification Certification Jump to: navigation, search Name LEED Certification/U.S. Green Building Council Address U.S. Green Building Council 2101 L Street, NW Suite 500 Place Washington, District of Columbia Zip 20037 Sector Buildings Product Certification system Website http://www.usgbc.org/DisplayPa Coordinates 38.889471°, -77.0578855° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":38.889471,"lon":-77.0578855,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

383

New Reactor Designs  

U.S. Energy Information Administration (EIA)

LWRs generate power through steam turbines similar to those used for most power generated by burning coal or fuel ... combined with British Nuclear Fuels Limited to ...

384

Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars  

SciTech Connect

The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with {approx}0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

Fensin, Michael L. [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL 32611 (United States); Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Elliott, John O. [Jet Propulsion Laboratories, California Institute of Technology, Pasedena, Ca 91109 (United States); Lipinski, Ronald J. [Sandia National Laboratory, Albuquerque, NM 87185 (United States); Poston, David I. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2006-01-20T23:59:59.000Z

385

The design of a functionally graded composite for service in high temperature lead and lead-bismuth cooled nuclear reactors  

E-Print Network (OSTI)

A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required ...

Short, Michael Philip

2010-01-01T23:59:59.000Z

386

Thermal hydraulic design of a 2400 MW t?h? direct supercritical CO?-cooled fast reactor  

E-Print Network (OSTI)

The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled ...

Pope, Michael A. (Michael Alexander)

2006-01-01T23:59:59.000Z

387

Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation  

E-Print Network (OSTI)

The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

Buongiorno, J.

388

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors  

E-Print Network (OSTI)

K. T. Assessment of Candidate Molten Salt Coolants for theK. T. Assessment of Candidate Molten Salt Coolants for thebeginning efforts for a molten salt reactor (MSR) program.

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

389

Conceptual design of a lead-bismuth cooled fast reactor with in-vessel direct-contact steam generation  

E-Print Network (OSTI)

The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

Buongiorno, Jacopo, 1971-

2001-01-01T23:59:59.000Z

390

Application of USNRC NUREG/CR-6661 and draft DG-1108 to evolutionary and advanced reactor designs  

Science Conference Proceedings (OSTI)

For the seismic design of evolutionary and advanced nuclear reactor power plants, there are definite financial advantages in the application of USNRC NUREG/CR-6661 and draft Regulatory Guide DG-1108. NUREG/CR-6661, 'Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems', was by Brookhaven National Laboratory (BNL) for the USNRC, and Draft Regulatory Guide DG-1108 is the proposed revision to the current Regulatory Guide (RG) 1.92, Revision 1, 'Combining Modal Responses and Spatial Components in Seismic Response Analysis'. The draft Regulatory Guide DG-1108 is available at http://members.cox.net/apolloconsulting, which also provides a link to the USNRC ADAMS site to search for NUREG/CR-6661 in text file or image file. The draft Regulatory Guide DG-1108 removes unnecessary conservatism in the modal combinations for closely spaced modes in seismic response spectrum analysis. Its application will be very helpful in coupled seismic analysis for structures and heavy equipment to reduce seismic responses and in piping system seismic design. In the NUREG/CR-6661 benchmark program, which investigated coupled seismic analysis of structures and equipment or piping systems with different damping values, three of the four participants applied the complex mode solution method to handle different damping values for structures, equipment, and piping systems. The fourth participant applied the classical normal mode method with equivalent weighted damping values to handle differences in structural, equipment, and piping system damping values. Coupled analysis will reduce the equipment responses when equipment, or piping system and structure are in or close to resonance. However, this reduction in responses occurs only if the realistic DG-1108 modal response combination method is applied, because closely spaced modes will be produced when structure and equipment or piping systems are in or close to resonance. Otherwise, the conservatism in the current Regulatory Guide 1.92, Revision 1, will overshadow the advantage of coupled analysis. All four participants applied the realistic modal combination method of DG-1108. Consequently, more realistic and reduced responses were obtained. (authors)

Chang 'Apollo', Chen [Apollo Consulting, Inc., Surprise, AZ 85374-4605 (United States)

2006-07-01T23:59:59.000Z

391

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

392

Document driven certification of computational science and engineering software  

Science Conference Proceedings (OSTI)

This paper presents a documentation and development methodology to facilitate the certification of Computational Science and Engineering (CSE) software that is produced by professional end user developers to solve mathematical models of physical systems. ... Keywords: computational science and engineering, document driven design, literate programming, software engineering, software requirements specification

Spencer Smith, Nirmitha Koothoor, Nedialko Nedialkov

2013-11-01T23:59:59.000Z

393

Earn ENERGY STAR certification | ENERGY STAR Buildings & Plants  

NLE Websites -- All DOE Office Websites (Extended Search)

Earn ENERGY STAR certification Earn ENERGY STAR certification Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In this section How can we help you? Build an energy program Improve building and plant performance Earn the ENERGY STAR and other recognition Benchmark energy use ENERGY STAR in action Communicate and educate ENERGY STAR communications toolkit Bring Your Green to Work with ENERGY STAR

394

DataTrends: ENERGY STAR Certification | ENERGY STAR Buildings & Plants  

NLE Websites -- All DOE Office Websites (Extended Search)

ENERGY STAR Certification ENERGY STAR Certification Secondary menu About us Press room Contact Us Portfolio Manager Login Facility owners and managers Existing buildings Commercial new construction Industrial energy management Small business Service providers Service and product providers Verify applications for ENERGY STAR certification Design commercial buildings Energy efficiency program administrators Commercial and industrial program sponsors Associations State and local governments Federal agencies Tools and resources Training In This Section Campaigns Commercial building design Communications resources Energy management guidance Financial resources Portfolio Manager Products and purchasing Recognition Research and reports Service and product provider (SPP) resources Success stories Target Finder

395

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01T23:59:59.000Z

396

Summary Protocol: Identification, Characterization, Designation, Remedial  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Summary Protocol: Identification, Characterization, Designation, Summary Protocol: Identification, Characterization, Designation, Remedial Action, Certification Summary Protocol: Identification, Characterization, Designation, Remedial Action, Certification Summary Protocol: Identification, Characterization, Designation, Remedial Action, Certification (January 1986) Summary Protocol: Identification, Characterization, Designation, Remedial Action, Certification (January 1986) More Documents & Publications Supplement No. 1 to the FUSRAP Summary Protocol - Designation/Elimination Protocol Pre-MARSSIM Surveys in a MARSSIM World: Demonstrating How Pre-MARSSIM Radiological Data Demonstrate Protectiveness at Formerly Utilized Sites Remedial Action Program Sites U.S. Department of Energy Guidelines for Residual Radioactive Material at

397

Considerations for Emerging Markets for Energy Savings Certificates  

E-Print Network (OSTI)

certificates; ESCs; energy efficiency; renewable energy;renewable energy certificates; RECs;state policy; renewable energy development; state policy

Friedman, Barry

2009-01-01T23:59:59.000Z

398

Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)  

DOE Green Energy (OSTI)

This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Mustafa Sacit [ORNL

2011-02-01T23:59:59.000Z

399

Grant County PUD - EnergyStar New Home Certification Program | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EnergyStar New Home Certification Program EnergyStar New Home Certification Program Grant County PUD - EnergyStar New Home Certification Program < Back Eligibility Construction Residential Savings Category Appliances & Electronics Heating & Cooling Home Weatherization Construction Commercial Weatherization Commercial Heating & Cooling Design & Remodeling Heat Pumps Program Info State District of Columbia Program Type Utility Rebate Program Rebate Amount Energy Star Manufactured Homes: $500 Site-Built Energy Star NW Homes with Heat Pump: $2,100 Heat Pumps: $200/ton Clothes Washers: $70 Provider Grant County PUD Grant County PUD offers incentives to qualifying customers through the Energy Star Home Certification Program. The Energy Star certification is available for both new site-built homes and new manufactured homes.

400

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009  

Science Conference Proceedings (OSTI)

This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

2010-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor design certification" 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

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008  

Science Conference Proceedings (OSTI)

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01T23:59:59.000Z

402

Official Certificate List Level(s)* Academic Administrative  

E-Print Network (OSTI)

in Educational Policy Studies EDUU S/295 Certificate in Energy Analysis and Policy at the Graduate Level IESG310&SU101 Certificate in Air Resources Management IESG104 Certificate in American Indian Studies L&SU S/110 Certificate in Archaeology L&SU S/120 Certificate in Asian American Studies L&SU201 Certificate

Sheridan, Jennifer

403

GRISET certification package  

Science Conference Proceedings (OSTI)

The GRISET (GRIMHX setup) computer code is an input driver for the three dimensional G computer code (Reference 4). GRISET provides a method for automating the creation of G geometry and associated problem specific information such as the reactor maps and the cross section data references. GRISET provides methods which allow the user to model control rod positioning, safety rod insertion with any number failing, control rod withdrawal or insertion, and full or partial melting of control rods. GRISET also calculates the radial shape factor, tilt, roof-top-ratios, and the assembly power peaking.

Taylor, J.J.; Cefus, G.R.

1992-11-01T23:59:59.000Z

404

THE HGCR-1, A DESIGN STUDY OF A NUCLEAR POWER STATION EMPLOYING A HIGH- TEMPERATURE GAS-COOLED REACTOR WITH GRAPHITE-UO$sub 2$ FUEL ELEMENTS  

SciTech Connect

The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)

Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J.; Samuels, G.

1959-07-28T23:59:59.000Z

405

Design of an Actinide Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low Cost Electricity FY-01 Annual Report, October 2001  

SciTech Connect

The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from work on this project (since project inception) are listed in Appendix A.

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Herring, James Stephen; Loewen, Eric Paul; Smolik, Galen Richard; Weaver, Kevan Dean; Todreas, N.

2001-10-01T23:59:59.000Z

406

Design of an Actinide Burning, Lead or Lead-Bismuth Cooled Reactor That Produces Low Cost Electricty - FY-02 Annual Report  

SciTech Connect

The purpose of this collaborative Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) Laboratory Directed Research and Development (LDRD) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, plant engineering, material compatibility studies, and coolant activation. The publications derived from work on this project (since project inception) are listed in Appendix A. This is the third in a series of Annual Reports for this project, the others are also listed in Appendix A as FY-00 and FY-01 Annual Reports.

Mac Donald, Philip Elsworth; Buongiorno, Jacopo

2002-10-01T23:59:59.000Z

407

SMART GRID TECHNOLOGIES (CERTIFICATE) www.ee.washington.edu/admissions/pmp/certificates  

E-Print Network (OSTI)

SMART GRID TECHNOLOGIES (CERTIFICATE) www.ee.washington.edu/admissions/pmp/certificates NOT ADMITTING FOR FALL 2011 ~ NEXT OFFERING OF SMART GRID CERTIFICATE TENTATIVELY FALL 2012 The Certificate in Smart Grid Technologies & Management in Electrical Engineering serves the growing need for engineers

Washington at Seattle, University of

408

Microsoft Word - CERTIFICATION LEVEL REQUIREMENTS.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CERTIFICATION LEVEL CERTIFICATION LEVEL The CEG is intended to provide program secretarial officers and field element managers (including operations offices, site offices, area offices, project offices, and service centers) with the requirements and guidelines for evaluating PMCDP candidate competencies and requests for equivalencies at all four certification levels and continuing education. Persons planning to be certified under the PMCDP may attain certification levels with the following total project cost (TPC) limits: * Certification Level 4: TPC exceeding $400 million (M) * Certification Level 3: TPC greater than $100M and equal to or less than $400M * Certification Level 2: TPC greater than $20M and equal to or less than $100M * Certification Level 1: TPC greater than $5M and equal to or less than $20M

409

PREPARING FOR A SUCCESSFUL EVMS CERTIFICATION  

Science Conference Proceedings (OSTI)

The client, a government agency, requires its contractor to obtain an Earned Value Management System (EVMS) certification that meets the intent of ANSI/EIA-748-B, Earned Value Management Systems. The contractor has extensive experience with certification preparation, having completed two certifications within two years. Information from a previous EVMS certification and internal system surveillances are used to prepare for client-based EVMS certifications and bi-annual surveillances. The contractor also sent members of its group to assist other companies preparing for surveillances and certifications to perform 'Black Hat Reviews.' This paper is a lessons learned on preparing a team for EVMS certification. The information is also applicable for surveillances, since the contractor prepares its team for the surveillance in the same manner as the initial certification. Some of the areas covered include required documents, tracing the data through the systems, Control Account Manager (CAM) preparation, and system verification.

CROWE SL; BASCHE AD

2011-02-09T23:59:59.000Z

410

Framework for Certification of Fish Propagation, Protection and Monitoring Facilities. Final Report.  

DOE Green Energy (OSTI)

A conceptual framework for certification of fish production and monitoring facilities including software templates to expedite implementation of the framework are presented. The framework is based on well established and widely utilized project management techniques. The implementation templates are overlays for Microsoft Professional Office software products: Excel, Word, and Project. Use of the software templates requires Microsoft Professional Office. The certification framework integrates two classical project management processes with a third process for facility certification. These processes are: (1) organization and definition of the project, (2) acquisition and organization of project documentation, and (3) facility certification. The certification process consists of systematic review of the production processes and the characteristics of the produced product. The criteria for certification review are the plans and specifications for the products and production processes that guided development of the facility. The facility is certified when the production processes are operating as designed and the product produced meets specifications. Within this framework, certification is a performance based process, not dissimilar from that practiced in many professions and required for many process, or a product meets professional/industry standards of performance. In the case of fish production facilities, the certifying authority may be diffuse, consisting of many entities acting through a process such as NEPA. A cornerstone of certification is accountability, over the long term, for the operation and products of a facility. This is particularly important for fish production facilities where the overall goal of the facility may require decades to accomplish.

Carlson, Thomas J.; Costello, Ronald J.

1997-06-01T23:59:59.000Z

411

NIF's Operations Support Building awarded LEED certification...  

NLE Websites -- All DOE Office Websites (Extended Search)

NIF's Operations Support Building awarded LEED certification | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the...

412

certification, compliance and enforcement regulations for Commercial...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

needs to be redone. certification, compliance and enforcement regulations for Commercial Refrigeration Equipment (CRE) More Documents & Publications Regulatory Burden RFI...

413

Building Energy Performance Certificate. Asset Rating.  

U.S. Energy Information Administration (EIA)

Non-Domestic Building Energy Performance Asset Rating ... Asset Rating. Author: BRE Subject: LCEA029636 Keywords: Energy Performance Certificate ...

414

Certification testing for small wind turbines  

DOE Green Energy (OSTI)

This paper describes the testing procedures for obtaining type certification for a small wind turbine. Southwest Windpower (SWWP) is seeking type certification from Underwriters Laboratory (UL) for the AIR 403 wind turbine. UL is the certification body and the National Renewable Energy Laboratory (NREL) is providing technical assistance including conducting the certification testing. This is the first small turbine to be certified in the US, therefore standards must be interpreted and test procedures developed.

Corbus, D.; Link, H.; Butterfield, S.; Stork, C.; Newcomb, C.

1999-10-20T23:59:59.000Z

415

Current Projects for Reactor Physics and Fuel Cycle Analysis...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Systems Modeling and Design Analysis > Reactor Physics and Fuel Cycle Analysis > Current Projects Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics...

416

THE ORNL GCR-3, A 750-Mw(e) GAS-COOLED CLAD-FUEL REACTOR POWER PLANT. A JOINT DESIGN STUDY  

SciTech Connect

ABS>An advanced, gas-cooled, clad-fuel reactor power plant to generate 750 Mw of electricity was designed as a study of the potential capability of that system. The graphitemoderated reactor generates 1908 Mw of heat in 1062 fuel channels 21 ft long for a power density of 5.5 kw/liter. Gas temperatures entering and leaving the reactor are 574 and 1150 deg F, respectively, operating at 420 psia. Steam at 2415 psia and 950 deg F with reheat to 1000 deg F drives a 763-Mw(e) turbogenerator and also four 31,000-hp blower drive turbines and the boiler feed pumps. Net thermal efficiency of the plant is 39.4%. Estimated direct cost of construction is 0,267,000, or 7 per kilowatt net electric output. Fuel-cycle costs at 20,000 Mwd per metric ton of uranium are 1.46 mills/ kwhr, operating and maintenance costs are 0.39 mill, and fixed charges range from 1.80 to 4.65 mills, depending on method of financing. Total power generation costs at an 80% load factor range from 3.65 to 6.50 mills/kwhr. (auth)

1963-02-01T23:59:59.000Z

417

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors  

SciTech Connect

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

2013-11-29T23:59:59.000Z

418

DOE Certification Programs | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE Certification Programs DOE Certification Programs DOE Certification Programs The following are certification programs offered at DOE: Acquisition Career Management Program (ACMP) - is managed by the Office of Acquisition and Project Management's (OAPM) Professional Development Division (PDD). ACMP implements Federal Acquisition Certifications established by the Office of Federal Procurement Policy in accordance with the provisions of 41 USC 433 regarding professionalization of the non-DOD acquisition workforce. IT Project Management Qualification Process - is developed by the Office of the Chief Information Officer (OCIO) to provide a framework for certifying Federal IT Project Managers as having the qualifications necessary to manage IT projects. Per the Office of Management and Budget

419

Equipment Certification Requirements | Open Energy Information  

Open Energy Info (EERE)

Equipment Certification Requirements Equipment Certification Requirements Jump to: navigation, search Policies requiring renewable energy equipment to meet certain standards serve to protect consumers from buying inferior equipment. These requirements not only benefit consumers; they also protect the renewable energy industry by making it more difficult for substandard systems to reach the market. [1] Contents 1 Equipment Certification Incentives 2 References Equipment Certification Incentives CSV (rows 1 - 19) Incentive Incentive Type Place Applicable Sector Eligible Technologies Active Canada Oil and Gas Operations Act (Canada) Environmental Regulations Equipment Certification Fees Generating Facility Rate-Making Generation Disclosure Industry Recruitment/Support Safety and Operational Guidelines

420

FAQS Qualification Card - NNSA Package Certification Engineer |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NNSA Package Certification Engineer NNSA Package Certification Engineer FAQS Qualification Card - NNSA Package Certification Engineer A key element for the Department's Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA). For each functional area, the FAQS identify the minimum technical competencies and supporting knowledge and skills for a typical qualified individual working in the area. FAQC-NNSAPackageCertificationEngineer.docx Description NNSA Package Certification Engineer Qualification Card

Note: This page contains sample records for the topic "reactor design certification" 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

Compliance Certification Enforcement | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Compliance Certification Enforcement Compliance Certification Enforcement Compliance Certification Enforcement DOE enforces the energy and water conservation regulations established by the Office of Energy Efficiency and Renewable Energy, Appliance Standards Program to ensure that residential, commercial and industrial equipment deliver the energy and cost savings required by law. To ensure that all covered products comply with the energy and water conservation standards, the enforcement office reviews manufacturers' compliance with certification requirements. Products covered under the Energy Policy and Conservation Act of 1975, as amended, are required annually to ''certify by means of a certification report that each basic model(s) meets the applicable energy conservation standard." This program

422

Offline count-limited certificates  

Science Conference Proceedings (OSTI)

In this paper, we present the idea of offline count-limited certificates (or clics for short), and show how these can be implemented using minimal trusted hardware functionality already widely available today. Offline count-limited ... Keywords: authentication, offline payments, smartcards, trusted platform module (TPM)

Luis F. G. Sarmenta; Marten van Dijk; Jonathan Rhodes; Srinivas Devadas

2008-03-01T23:59:59.000Z

423

Postgraduate Certificate in Geothermal Energy  

E-Print Network (OSTI)

Postgraduate Certificate in Geothermal Energy Technology Department of Engineering Science to study for the PGCertGeothermTech will require a visa. Details about how to obtain a visa to study in New your visa. Geothermal Training in New Zealand New Zealand is a beautiful country in the South Pacific

Auckland, University of

424

BOILING WATER REACTOR TRANSIENT INSTABILITY STUDIES OF RINGHALS 1 REACTOR USING TRACE COUPLED WITH PARCS.  

E-Print Network (OSTI)

??Reactor plant design often incorporates data and insight ascertained from computer code simulations of plant dynamics and reactor core behavior. Increasing utilization of data gathered… (more)

Walls, Robert

2009-01-01T23:59:59.000Z

425

Directory of Certificates of Compliance for radioactive materials packages: Certificates of Compliance. Volume 2, Revision 18  

Science Conference Proceedings (OSTI)

The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 of approved QA programs. The reports include a listing of all users of each package design and approved QA programs prior to the publication date.

NONE

1995-10-01T23:59:59.000Z

426

Design of slurry bubble column reactors: novel technique for optimum catalyst size selection contractual origin of the invention  

DOE Patents (OSTI)

A method for determining optimum catalyst particle size for a gas-solid, liquid-solid, or gas-liquid-solid fluidized bed reactor such as a slurry bubble column reactor (SBCR) for converting synthesis gas into liquid fuels considers the complete granular temperature balance based on the kinetic theory of granular flow, the effect of a volumetric mass transfer coefficient between the liquid and the gas, and the water gas shift reaction. The granular temperature of the catalyst particles representing the kinetic energy of the catalyst particles is measured and the volumetric mass transfer coefficient between the gas and liquid phases is calculated using the granular temperature. Catalyst particle size is varied from 20 .mu.m to 120 .mu.m and a maximum mass transfer coefficient corresponding to optimum liquid hydrocarbon fuel production is determined. Optimum catalyst particle size for maximum methanol production in a SBCR was determined to be in the range of 60-70 .mu.m.

Gamwo, Isaac K. (Murrysville, PA); Gidaspow, Dimitri (Northbrook, IL); Jung, Jonghwun (Naperville, IL)

2009-11-17T23:59:59.000Z

427

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Treshow, M.

1958-11-18T23:59:59.000Z

428

Materials Reliability Program: Sensitivity Studies for Functionality Analysis of Reactor Internals of Westinghouse-Designed Power Plants (MRP-356)  

Science Conference Proceedings (OSTI)

In developing the Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0), a semi-empirical material behavior model for irradiated austenitic stainless steels was developed for the degradation mechanisms applicable to the internals. Functionality analyses predicting aging up to 60 years were performed using the material model. MRP-230 results indicate that susceptibility to irradiation-assisted stress corrosion cracking (IASCC) is governed by two competing ...

2013-08-13T23:59:59.000Z

429

PRELIMINARY DESIGN AND COST ESTIMATE FOR THE PRODUCTION OF CENTRAL STATION POWER FROM AN AQUEOUS HOMOGENEOUS REACTOR UTILIZING THORIUM-URANIUM-233  

SciTech Connect

The design and economics of the Aqueous Homogeneous Reactor as basically under development at the Oak Ridge National Laboratory are presented. The reactor system utilizes thorium-U-233 fuel. Conditions accompanying reactor systems generating up to l080 mw of net electrical energy are covered. The study indicates that a generating station, with a net thermal efficiency of 28.l%, might be constructed for approximately 0/kw and 0/kw at the l80 mw and l080 mw electrical levels, respectively. These values result in capital expenses of approximately 4.72 and 2.86 milis/kwh. A major part of fuel cost is the expense of chemical processing. It is therefore advantageous 10 schedule fuel through a relatively large processing system since fixed charges are insensitive to chemical plant size. By handling fuel through a plant large enough for processing 200 kg of thorium per day, total fuel costa of about 1 mill/kwh result. This cost for fuel processing appears applicable to generating stations up to abeut 540 mw in size, decreasing to about 0.6 mills/kwh at the l080 mw level. Operating and maintenance expense, including heavy water cost on a lease basis, varies between l.34 and 0.89 mills/kwh for l80 and l080 megawatts respectively. If the purchase of heavy water is required, 0.3 to 0.4 mills/kwh must be added. It is concluded that the Aqueous Homogeneous Reactor may produce electrical power competitive with conventional generating systems when the remaining technical problems are solved. It is felt ihat the research and development now programed by the Oak Ridge National Laboratory will solve these problems and affect costs favorably. (auth)

Carson, H.G.; Landrum, L.H. eds.

1955-02-01T23:59:59.000Z

430

CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues  

Science Conference Proceedings (OSTI)

Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule ({section}50.62); (2) station blackout ({section}50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term ({section}50.34(f) and {section}100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements ({section}50.55a, etc); (6) ECCS acceptance criteria ({section}50.46)(b); (7) combustible gas control ({section}50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle ({section}51.51); and (11) (standards {section}50.55a).

Charak, I.; Kier, P.H. [Argonne National Lab., IL (United States)

1995-04-01T23:59:59.000Z

431

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first ref