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1

Recovery Act's HWCTR Project Empty of Equipment, Ready for Grouting...  

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

from the American Recovery and Reinvestment Act, the next phase has begun on decommissioning the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site (SRS)....

2

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

Science Conference Proceedings (OSTI)

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

Vinson, Dennis

2010-06-01T23:59:59.000Z

3

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

Science Conference Proceedings (OSTI)

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

Austin, W.; Brinkley, D.

2011-10-13T23:59:59.000Z

4

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

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

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

5

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

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

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

6

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project |  

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

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project February 1, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE (803) 952-7697 james-r.giusti@srs.gov Paivi Nettamo, SRNS (803) 646-6075 paivi.nettamo@srs.gov AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and Reinvestment Act. Affectionately known by SRS employees as "Hector," the iconic Heavy Water Components Test Reactor (HWCTR) has stood in the Site's B Area since 1959

7

--No Title--  

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

(HWCTR), located in B-Area. These activities are being conducted prior to the decommissioning of 770-U, HWCTR. The HWCTR Facility was a test reactor that tested fuel...

8

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

9

HEAVY WATER MODERATED POWER REACTORS. Progress Report for October 1959  

SciTech Connect

Continued progress is reported on the design and construction of the Heavy Water Components Test Reactor; 78% of the firm design and 17% of the construction were complete at the end of October 1959. Approximateiy 15% of the firm design for the isolated coolant loops of the HWCTR was also complete. The results of further fabrication tests and irradiation tests of fuel tubes of natural uranium metal are reported. One of the metal tubes failed under irradiation, while other irradiations of metal fuels progressed satisfactorily. (auth)

Hood, R.R.; Isakoff, L. comps.

1959-11-01T23:59:59.000Z

10

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1  

SciTech Connect

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Owen, M.B.

1997-04-01T23:59:59.000Z

11

Idaho National Laboratory Advanced Test Reactor Probabilistic...  

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

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012...

12

Microstructural Characterization of Test Reactor Irradiated RPV ...  

Science Conference Proceedings (OSTI)

Presentation Title, Microstructural Characterization of Test Reactor Irradiated RPV ... Evolution in High Purity Reference V-4Cr-4Ti Alloy for Fusion Reactor.

13

Material Science Advances Using Test Reactor Facilities  

Science Conference Proceedings (OSTI)

Aug 2, 2010 ... About this Symposium. Meeting, 2011 TMS Annual Meeting & Exhibition. Symposium, Material Science Advances Using Test Reactor Facilities.

14

PIA - Advanced Test Reactor National Scientific User Facility...  

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

Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

15

Ground test facility for nuclear testing of space reactor subsystems  

SciTech Connect

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs.

Quapp, W.J.; Watts, K.D.

1985-01-01T23:59:59.000Z

16

REACTOR FUEL ELEMENTS TESTING CONTAINER  

DOE Patents (OSTI)

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Whitham, G.K.; Smith, R.R.

1963-01-15T23:59:59.000Z

17

TEST REACTORS MEETING FOR INDUSTRY, IDAHO FALLS, IDAHO, MAY 13-15, 1959. PART I. CONSTRUCTION AND OPERATION OF TEST REACTORS. PART II. UTILIZATION OF TEST REACTORS  

SciTech Connect

Twelve papers on construction and operation of test reactors and nine papers on the utilization of test reactors are presented.(W.D.M.)

1959-10-31T23:59:59.000Z

18

CX-000509: Categorical Exclusion Determination | Department of...  

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

9: Categorical Exclusion Determination CX-000509: Categorical Exclusion Determination Heavy Water Component Test Reactor (HWCTR) Well Installation CX(s) Applied: B3.1 Date: 0805...

19

Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

20

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...

Note: This page contains sample records for the topic "test reactor hwctr" 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

Mechanical Testing of Core Fast Reactor Materials for the Advanced ...  

Science Conference Proceedings (OSTI)

To achieve this goal, the core fast reactor materials (cladding and duct) must be ... in situ Mechanical Test Methods in the US Fusion Reactor Materials Program.

22

 

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

The proposed activity is the Deactivation of the 770-U, Heavy Water Components Test Reactor (HWCTR), located in B-Area. These activities are The proposed activity is the Deactivation of the 770-U, Heavy Water Components Test Reactor (HWCTR), located in B-Area. These activities are being conducted prior to the decommissioning of 770-U, HWCTR. The HWCTR Facility was a test reactor that tested fuel components for heavy water reactors. The reactor was deactivated in 1965 and had an asbestos abatement program in effect in 1994 where essentially all of the asbestos was removed. The majority of the radioactivity is in the reactor internals and shielding as activation products. Deactivation of Heavy Water Components Test Reactor (HWCTR), 770-U Savannah River Site Aiken South Carolina ARRA - U - 2010 - 014, Rev.0 23Feb10 Andrew R. Grainger Digitally signed by Andrew R. Grainger DN: CN = Andrew R. Grainger, C = US, O =

23

Final safeguards analysis, High Temperature Lattice Test Reactor  

SciTech Connect

Information on the HTLTR Reactor is presented concerning: reactor site; reactor buildings; reactor kinetics and design characteristics; experimental and test facilitles; instrumentation and control; maintenance and modification; initial tests and operations; administration and procedural safeguards; accident analysis; seifterminated excursions; main heat exchanger leak; training program outline; and reliability analysis of safety systems. (7 references) (DCC)

Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

1966-01-01T23:59:59.000Z

24

Instrumentation to Enhance Advanced Test Reactor Irradiations  

SciTech Connect

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

2009-09-01T23:59:59.000Z

25

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

26

Decommissioning of the Tokamak Fusion Test Reactor  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

2003-10-28T23:59:59.000Z

27

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 1. REACTOR SCIENCE AND TECHNOLOGY  

SciTech Connect

A resume of nuclear physics basic to reactor operation precedes discussion of aspects of reactor physics, engineering, chemistry, metallurgy, instrumentation, control, kinetics, and safety. The object is to provide an approach to and understanding of problems in irradiation test programs in the Materials Testing and Engineering Test Reactors. (D.C.W.)

1963-06-01T23:59:59.000Z

28

Nuclear reactor containment spray testing system. [PWR  

SciTech Connect

Disclosed is a method for periodic testing of a spray system in a nuclear reactor containment. The method includes injecting a gas into the spray system such that a temperature differential exists between the gas and the containment atmosphere. Scanning the gas jet discharged from the spray nozzles with infrared apparatus then provides a real-time thermal image on a monitor, such as a cathode ray tube, and detects any partially or completely blocked nozzles in the spray system. The scanning may be performed from the containment operating deck. 1 claim, 4 figures.

Rubin, K.

1978-01-10T23:59:59.000Z

29

Massive Hanford Test Reactor Removed- Plutonium Recycle Test Reactor removed from Hanford’s 300 Area  

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

RICHLAND, WA – Hanford’s River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy’s (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area.

30

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

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

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

31

The Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

Symposium, Materials Solutions for the Nuclear Renaissance ... U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific ...

32

Engineering Test Reactor (ETR) Vessel Relocated after 50 years.  

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

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

33

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

34

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

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

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

35

Space reactor fuel element testing in upgraded TREAT  

DOE Green Energy (OSTI)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-05-01T23:59:59.000Z

36

Space reactor fuel element testing in upgraded TREAT  

DOE Green Energy (OSTI)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-14T23:59:59.000Z

37

Testing of Biomass in a Transport Reactor Gasifier  

Science Conference Proceedings (OSTI)

A 200-hour gasification test was undertaken on biomass fuels from sources that include wood waste and a potential energy crop such as switchgrass. The test involved the design and construction of a feed system to allow 100% biomass to be continuously fed to the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center. Biomass performance was also assessed in a high-efficiency transport reactor gasifier, the centerpiece of an advanced biomass integrated ...

2012-11-28T23:59:59.000Z

38

Production test IP-412-AI: B and C reactors export system test  

SciTech Connect

Purpose of this test was to determine the adequacy of the export system for supplying flow to a dual reactor area under simulated emergency conditions.

Benson, J.L.; Jones, S.S.

1961-08-02T23:59:59.000Z

39

EBR-2 (Experimental Breeder Reactor-2) test programs  

SciTech Connect

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA)); Planchon, H.P.; Lambert, J.D.B.; Hill, D.J. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

40

EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs  

SciTech Connect

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

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

42

THERMAL PERFORMANCE OF A FAST NEUTRON TEST CONCEPT FOR THE ADVANCED TEST REACTOR  

Science Conference Proceedings (OSTI)

Since 1967, the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) has provided state-of-the-art experimental irradiation testing capability. A unique design is investigated herein for the purpose of providing a fast neutron flux test capability in the ATR. This new test capability could be brought on line in approximately 5 or 6 years, much sooner than a new test reactor could be built, to provide an interim fast-flux test capability in the timeframe before a fast-flux research reactor could be built. The proposed cost for this system is approximately $63M, much less than the cost of a new fast-flux test reactor. A concept has been developed to filter out a large portion of the thermal flux component by using a thermally conductive neutron absorber block. The objective of this study is to determine the feasibility of this experiment cooling concept.

Donna Post Guillen

2008-06-01T23:59:59.000Z

43

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC)

44

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

45

In-Situ Creep Testing Capability for the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2012-09-01T23:59:59.000Z

46

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR  

DOE Patents (OSTI)

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17T23:59:59.000Z

47

--No Title--  

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

Three monitoring wells will be installed around the Heavy Water Component Test Reactor (HWCTR, Building 770-U) located in B-Area. The wells will be 135' deep and 2" diameter. The...

48

--No Title--  

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

Four groundwater monitoring wells are to be installed around the Heavy Water Component Test Reactor (HWCTR) in B Area. The wells will be screened at 135ft in the water table...

49

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

50

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

51

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

52

Reduced enrichment for research and test reactors: Proceedings  

SciTech Connect

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

1988-05-01T23:59:59.000Z

53

Enhanced In-Pile Instrumentation at the Advanced Test Reactor  

SciTech Connect

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

2012-08-01T23:59:59.000Z

54

Enhanced In-Pile Instrumentation at the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; K. Condie

2011-06-01T23:59:59.000Z

55

The Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are shown in Figure 1. This article describes the ATR as well as the expanded capabilities, partnerships, and services that allow researchers to take full advantage of this national resource.

Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

56

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

2012-03-01T23:59:59.000Z

57

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

58

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs  

SciTech Connect

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

2011-09-01T23:59:59.000Z

59

DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR  

Science Conference Proceedings (OSTI)

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

Rule, Keith; Perry, Erik; Parsells, Robert

2003-02-27T23:59:59.000Z

60

Diamond Wire Cutting of the Tokamak Fusion Test Reactor  

Science Conference Proceedings (OSTI)

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

Keith Rule; Erik Perry; Robert Parsells

2003-01-31T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

Advanced Test Reactor National Scientific User Facility Progress  

SciTech Connect

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives cannot be met using the INL facilities. The ATR NSUF program includes a robust education program enabling students to participate in their research at INL and the partner facilities, attend the ATR NSUF annual User Week, and compete for prizes at sponsored conferences. Development of additional research capabilities is also a key component of the ATR NSUF Program; researchers are encouraged to propose research projects leading to these enhanced capabilities. Some ATR irradiation experiment projects irradiate more specimens than are tested, resulting in irradiated materials available for post irradiation examination by other researchers. These “extra” specimens comprise the ATR NSUF Sample Library. This presentation will highlight the ATR NSUF Sample Library and the process open to researchers who want to access these materials and how to propose research projects using them. This presentation will provide the current status of all the ATR NSUF Program elements. Many of these were not envisioned in 2007, when DOE established the ATR NSUF.

Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

2012-10-01T23:59:59.000Z

62

Test storage of spent reactor fuel in the Climax granite at the Nevada Test Site  

SciTech Connect

A test of retrievable dry geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.3 years out of reactor core (about 2 kW/canister thermal output) will be emplaced in a storage drift along with 6 electrical simulator canisters and their effects will be compared. Two adjacent drifts will contain electrical heaters, which will be operated to simulate within the test array the thermal field of a large repository. The test objectives, technical concepts and rationale, and details of the test are stated and discussed.

Ramspott, L.D.; Ballou, L.B.

1980-02-13T23:59:59.000Z

63

Technology issues for decommissioning the Tokamak Fusion Test Reactor  

SciTech Connect

The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community.

Spampinato, P.T.; Walton, G.R. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Commander, J.C. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1994-07-01T23:59:59.000Z

64

INITIAL TESTING AND OPERATION OF THE ARGONNE LOW POWER REACTOR (ALPR)  

SciTech Connect

The major events of a program designed to test and operate the completed reactor power plant and associated equipment are described. The design and construction phases of the project, component installation, preliminary systems testing, zero-power experiments, areas affected by the design parameters, reactor operation, plant safety, and reactor operator training are covered. (W.D.M.)

Hamer, E.E. ed.

1959-12-01T23:59:59.000Z

65

Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project  

SciTech Connect

This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

A. B. Culp

2007-01-26T23:59:59.000Z

66

2011_sr_03.doc  

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

Wednesday, February 23, 2011 Wednesday, February 23, 2011 james-r.giusti@srs.gov Paivi Nettamo, SRNS, (803) 646-6075 paivi.nettamo@srs.gov Recovery Act's HWCTR Project Empty of Equipment, Ready for Grouting AIKEN, S.C. - Thanks to investments from the American Recovery and Reinvestment Act, the next phase has begun on decommissioning the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site (SRS).

67

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor  

SciTech Connect

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

T. A. Tomberlin; S. B. Grover

2004-11-01T23:59:59.000Z

68

Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)  

E-Print Network (OSTI)

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

69

Advanced LWR Fuel Testing Capabilities in the ORNL High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

A new test capability for the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is being developed that will allow testing of advanced nuclear fuels and cladding materials under prototypic light-water reactor (LWR) operating conditions in less time than it takes in other research reactors. This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiments currently planned to start in late 2008.

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL; Spellman, Donald J [ORNL

2008-01-01T23:59:59.000Z

70

Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent operating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile are similar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.

G. S. Chang; R. G. Ambrosek

2005-11-01T23:59:59.000Z

71

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

72

Implementation of a testing and diagnostic concept for an NPP reactor protection system  

Science Conference Proceedings (OSTI)

This paper presents the concept and practical realization of the testing and diagnostic methodology for a reactor protection system in a nuclear power plant. The test concept utilizes the highly redundant nature of these systems to conduct tests during ...

Tamás Bartha; István Varga; Alexandros Soumelidis; Géza Szabé

2005-04-01T23:59:59.000Z

73

Testing of an advanced thermochemical conversion reactor system  

DOE Green Energy (OSTI)

This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

Not Available

1990-01-01T23:59:59.000Z

74

Enhanced In-pile Instrumentation for Material Testing Reactors  

Science Conference Proceedings (OSTI)

An increasing number of U.S. nuclear research programs are requesting enhanced in-pile instrumentation capable of providing real-time measurements of key parameters during irradiations. For example, fuel research and development funded by the U.S. Department of Energy now emphasize approaches that rely on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time data are essential for characterizing the performance of new fuels during irradiation testing. Furthermore, sensors that obtain such data must be miniature, reliable and able to withstand high flux/high temperature conditions. Depending on user requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these user needs, in-pile instrumentation development efforts have been initiated as part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF), the Fuel Cycle Research & Development (FCR&D), and the Nuclear Energy Enabling Technology (NEET) programs. This paper reports on recent INL achievements to support these programs. Specifically, an overview of the types of sensors currently available to support in-pile irradiations and those sensors currently available to MTR users are identified. In addition, recent results and products available from sensor research and development are detailed. Specifically, progress in deploying enhanced in-pile sensors for detecting elongation and thermal conductivity are reported. Results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are also summarized.

Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley

2012-07-01T23:59:59.000Z

75

High uranium density dispersion fuel for the reduced enrichment of research and test reactors program.  

E-Print Network (OSTI)

??This work describes the fabrication of a high uranium density fuel for the Reduced Enrichment of Research and Test Reactors Program. In an effort to… (more)

[No author

2006-01-01T23:59:59.000Z

76

Final Site-Specific Decommissioning Inspection Report for the University of Washington Research and Test Reactor  

SciTech Connect

Report of site-specific decommissioning in-process inspection activities at the University of Washington Research and Test Reactor Facility.

Sarah Roberts

2006-10-18T23:59:59.000Z

77

Fast Flux Test Reactor: Re-evaluation of the Department's Approach...  

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

Sites Power Marketing Administration Other Agencies You are here Home Fast Flux Test Reactor: Re-evaluation of the Department's Approach to Deactivation, Decontamination,...

78

Deterministic Modeling of the High Temperature Test Reactor  

SciTech Connect

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.

Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

2010-06-01T23:59:59.000Z

79

Overview of Component Testing Requirements for a Small Fluoride Salt-Cooled High Tempreature Reactor  

Science Conference Proceedings (OSTI)

This article summarizes the information necessary to provide reasonable assurance that components for a small fluoride salt-cooled high temperature reactor will meet their functional requirements. In support of the analysis of testing requirements, a simplified, conceptual description of the systems, structures, and components specific to this reactor class was developed. These reactor system elements were divided into major categories based on their functions: (1) reactor core systems, (2) heat transport system, (3) reactor auxiliary cooling system, and (4) instrumentation and controls system. An assessment of technical maturity for each element was made, and a gap analysis was performed to identify specific areas that require further testing. A prioritized list of the testing requirements was then developed. The prioritization was based on both the relative importance of the system to reactor viability, and performance and time requirements to perform the testing.

Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2010-01-01T23:59:59.000Z

80

MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES. Quarterly Report No. 3, July 1-September 30, 1963  

SciTech Connect

8 6 < platelets containing U/sub 3/O/sub 8/, UO/sub 2/, or UAl/sub 3/ in aluminum matrices were irradiated in the ETR at inltial surface temperatures of 180 deg C to burnups of 1 x 10/sup 21/ fiss/ cm/sup 3/. The high fuel loadings (approximately 35 wt% U/sup 235/) in UO/sub 2/ and U/sub 3/O/sub 8/ blistered under these conditions; the UAl/sub samples were still in good condition at the end of the test. Electrolyzed coatings on aluminum deteriorated badly under exposures of 3 to 5 x 10/sup 20/ n/cm/sup 2/ (>1Mev) in the ETR process water. The ARMF-1 regulating rod was repaired and digital regulating rod position readout instrumentation installed during an extended shutdown after more than two years of operation. Fission product transient curves extrapolated to about the same zero time reactivity value with initial data varying from 30 minutes to 6 hours. This limited considerably the probability that short-lived high cross section fission products exist. Under present ETR operating conditions the maximum decrease in effectiveness of a nickel absorber section as a result of burnup would be less than 20% in 20 years. Thus, burnup appears not to be a factor which limits its useful life. The preliminary analysis and flow charts for the Phillips General Purpose Monte Carlo Program for the IBM 7040 are nearing completion. Two reactor simulation devices were put into service in the Analog Computer Facility, a reactor kinetics simulator and seven transport lag simulation channels. Preliminary design of a Xe-135 simulator was completed. The fission cross section of Pu/sup 241/ was measured from 2 to 100 ev. Resolution of the linear electron accelerator used was sufficient to permit multilevel analysis of the neutron levels below 36 ev. Transmission measurements were obtained on a separated Pa/sup 233/ sample containing approximately 10 mg of Pa/sup 2/O/sub 5/. The energies of these resonances observed with the unseparated sample with their relative sizes are presented. Several experiments were conducted to determine the useful lifetime of solid state detectors under in- pile conditions of fission fragment bombardment. A single detector, using an external U/sup 235/ fission source was irradiated to approximately 3 x 10/sup 9/ total fission, at which point the fission fragment peaks were still well resolved and the signal pulses were sufficientiy large compared to noise level so that the latter could be effectively biased out. Averaged reduced partial differential scatiering cross sections for a powder Be sample were obtained. A Data Processing System for transient data was developed for use in the SPERT reactor complex. Data are recorded on an FM tape system and applied to a magnetic memory for temporary storage and from there to one or more of several readout devices. An Eight-Input Adapter and an Initial Delay Counter were developed to increase the utility of an existing time-of-flight analyzer. A Personnel Monitor ( Frisker'') is described, which approaches closely an ideal monitor for use with widely varying radiation backgrounds. Current feedback around an operational amplifier is used to provide a current source used to drive oscillograph galvanometers thereby extending the range of linear operation of the galvanometers. The work of placing a large telemetered radiological survey system in operation is described along with the description of a remote station simulator. Dynamic pressure tests of several commercial transducers are described together with the criteria established for suitability for their use in reactor transient studies. Rod drop deceleration times were measured on an ETR control rod; the test instrumentation is described. The 7090 version of PDQ-4 (20,000 mesh points) was converted and modified for operation on the 7040. The following reactor codes are also now in operation on the 7040: TEMPEST-II, GAM, FOG, ZUT, MIST, ULCER, and TOPIC. Being de-bugged are HEAT-I and IREKIN. In addition, the following programs for the 7040 were written and placed in operation: matrix inversion, ordinary differ

1964-02-15T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" from the National Library of EnergyBeta (NLEBeta).
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81

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR  

SciTech Connect

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

2010-10-01T23:59:59.000Z

82

Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition  

Science Conference Proceedings (OSTI)

An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S. [and others

1995-09-01T23:59:59.000Z

83

Testing mass-varying neutrinos with reactor experiments  

E-Print Network (OSTI)

We propose that reactor experiments could be used to constrain the environment dependence of neutrino mass and mixing parameters, which could be induced due to an acceleron coupling to matter fields. There are several short-baseline reactor experiment projects with different fractions of air and earth matter along the neutrino path. Moreover, the short baselines, in principle, allow the physical change of the material between source and detector. Hence, such experiments offer the possibility for a direct comparison of oscillations in air and matter. We demonstrate that for sin 2 (2?13) ? 0.04, two reactor experiments (one air, one matter) with baselines of at least 1.5 km can constrain any oscillation effect which is different in air and matter at the level of a few per cent. Furthermore, we find that using the same experiment while physically moving the material between source and detector improves systematics. PACS: 14.60.Pq

unknown authors

2005-01-01T23:59:59.000Z

84

MATERIALS TESTING REACTOR PROJECT. QUARTERLY REPORT FOR PERIOD ENDING MARCH 1, 1950  

SciTech Connect

Progress is reported in finaiizing basic design data for the Materials Testing Reactor. The major emphasis at ANL was on issurance of design reports on practically all phases of the MTR project outside the reactor face and low the first fioor level. Operation of the mock-up reacr at ORNL at 10 watts resulted in no major design changes. Topics discussed include the reactor building, wing, and reactor service building; canal and canal facilities; water systems; air exhaust systems; electrical power systems; effluent control; and shielding requirements. 11 drawings. (C.H.)

Huffman, J.R.

1958-10-31T23:59:59.000Z

85

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing  

Science Conference Proceedings (OSTI)

New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

D. L. Knudson; J. L. Rempe

2012-02-01T23:59:59.000Z

86

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing  

Science Conference Proceedings (OSTI)

New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

D. L. Knudson; J. L. Rempe

2012-02-01T23:59:59.000Z

87

Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition  

Science Conference Proceedings (OSTI)

Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W. [and others

1995-08-01T23:59:59.000Z

88

Evaluation and Test of Improved Fire Resistant Fluid Lubricants for Water Reactor Coolant Pump Motors, Volume 1: Fluid Evaluation, Bearing Model Tests, Motor Tests, and Fire Tests  

Science Conference Proceedings (OSTI)

Commercially available fire-resistant fluid lubricants were evaluated to determine their suitability for use in primary-system pump motors in nuclear reactors. Volume 1 describes the procedures and results of tests of lubrication properties; fire and radiation resistance; and thermal, oxidative, and hydrolytic stability.

1980-07-01T23:59:59.000Z

89

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components  

SciTech Connect

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01T23:59:59.000Z

90

Production test IP-338-A, Supp. A, DR-Reactor heat decay test at high outlet water temperatures  

SciTech Connect

This test is identical to the original except that it authorizes the performance of a trial reduction in reactor flow during a prior reactor shutdown. This trial flow reduction will be performed in the same manner as proposed for the actual test, with one exception. This is, that based upon the results of this preliminary test some changes in the timing of the different steps may be indicated. Such changes can readily be handled by making each step dependent upon the observed reactor outlet temperature during the test performance. The other significant change in the production test is the increase in the allowable bulk outlet temperature from Ti + 40 {plus_minus} 3{degrees}C{sup *}. This change is needed to obtain a reasonable extrapolation of the results of tests No. 1 and No.2 to 90{degrees}C, and is justified from a hazards standpoint by the excellent flow control achieved during test No. 1 and by the trial test that will be run prior to the performance of the actual test No. 2. Other aspects of the test basis and justification are presented in the original production test.

Jones, S.S.

1962-05-18T23:59:59.000Z

91

10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion  

SciTech Connect

The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors [HPRR]) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

2012-05-01T23:59:59.000Z

92

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight 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). 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 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 ten years to demonstrate and qualify new 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 the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

S. Blaine Grover

2009-09-01T23:59:59.000Z

93

Fuel subassembly leak test chamber for a nuclear reactor  

DOE Patents (OSTI)

A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

Divona, Charles J. (Santa Ana, CA)

1978-04-04T23:59:59.000Z

94

MODELING ASSUMPTIONS FOR THE ADVANCED TEST REACTOR FRESH FUEL SHIPPING CONTAINER  

SciTech Connect

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is currently licensed per 10 CFR 71 to transport a fresh fuel element for either the Advanced Test Reactor, the University of Missouri Research Reactor (MURR), or the Massachusetts Institute of Technology Research Reactor (MITR-II). During the licensing process, the Nuclear Regulatory Commission (NRC) raised a number of issues relating to the criticality analysis, namely (1) lack of a tolerance study on the fuel and packaging, (2) moderation conditions during normal conditions of transport (NCT), (3) treatment of minor hydrogenous packaging materials, and (4) treatment of potential fuel damage under hypothetical accident conditions (HAC). These concerns were adequately addressed by modifying the criticality analysis. A tolerance study was added for both the packaging and fuel elements, full-moderation was included in the NCT models, minor hydrogenous packaging materials were included, and fuel element damage was considered for the MURR and MITR-II fuel types.

Rick J. Migliore

2009-09-01T23:59:59.000Z

95

Solar test of an integrated sodium reflux heat pipe receiver/reactor for thermochemical energy transport  

DOE Green Energy (OSTI)

A chemical reactor for carbon dioxide reforming of methane was integrated into a sodium reflux heat pipe receiver and tested in the solar furnace of the Weizmann Institute of Science, Rehovot, Israel. The receiver/reactor was a heat pipe with seven tubes inside an evacuated metal box containing sodium. The catalyst, 0.5 wt% Rh on alumina, filled two of the tubes with the front surface of the box serving as the solar absorber. In operation, concentrated sunlight heated the front plate and vaporized sodium from a wire mesh wick attached to other side. Sodium vapor condensed on the reactor tubes, releasing latent heat and returning to the wick by gravity. The receiver system performed satisfactorily in many tests under varying flow conditions. The maximum power absorbed was 7.5 kW at temperatures above 800C. The feasibility of operating a heat pipe receiver/reactor under solar conditions was proven, and the advantages of reflux devices confirmed.

Diver, R.B.; Fish, J.D. (Sandia National Labs., Albuquerque, NM (United States)); Levitan, R.; Levy, M.; Meirovitch, E.; Rosin, H. (Weizmann Inst. of Science, Rehovot (Israel)); Paripatyadar, S.A.; Richardson, J.T. (Univ. of Houston, TX (United States))

1992-01-01T23:59:59.000Z

96

Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test  

Science Conference Proceedings (OSTI)

This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy.

Cowell, B.S.

1997-06-01T23:59:59.000Z

97

Isotope correlation studies relative to high enrichment test reactor fuels  

SciTech Connect

Several correlations of fission product isotopic ratios with atom percent fission and neutron flux, for highly enriched /sup 235/U fuel irradiated in two different water moderated thermal reactors, have been evaluated. In general, excellent correlations were indicated for samples irradiated in the same neutron spectrum; however, significant differences in the correlations were noted with the change in neutron spectrum. For highly enriched /sup 235/U fuel, the correlation of the isotopic ratio /sup 143/Nd//sup 145 +146/Nd with atom percent fission has wider applicability than the other fission product isotopic ratio evaluated. The /sup 137/Cs//sup 135/Cs atom ratio shows promise for correlation with neutron flux. Correlations involving heavy element ratios are very sensitive to the neutron spectrum.

Maeck, W.J.; Tromp, R.L.; Duce, F.A.; Emel, W.A.

1978-06-01T23:59:59.000Z

98

The Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations  

E-Print Network (OSTI)

:1. Our range of sensitivity is tuned to test the š¯ $ še solution of the atmospheric neutrino anomaly. 11 The Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations 94305 e Palo Verde Nuclear Generating Station,Tonopah AZ 85354 Our collaboration has installed a long

Piepke, Andreas G.

99

The Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations  

E-Print Network (OSTI)

\\Gamma3 eV 2 and sin 2 2\\Theta ! 0:1. Our range of sensitivity is tuned to test the š ¯ $ š e solutionThe Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations Presented 85287 S. Pittalwala, R. Wilferd, S. Young Palo Verde Nuclear Generating Station, Tonopah AZ 85354 Our

Piepke, Andreas G.

100

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information  

SciTech Connect

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" from the National Library of EnergyBeta (NLEBeta).
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101

The ORNL High Flux Isotope Reactor and New Advanced Fuel Testing Capabilities  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL

2011-01-01T23:59:59.000Z

102

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

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

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

103

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

Science Conference Proceedings (OSTI)

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

104

Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests  

Science Conference Proceedings (OSTI)

Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

2011-06-15T23:59:59.000Z

105

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Ott, Larry J [ORNL; Ellis, Ronald James [ORNL; McDuffee, Joel Lee [ORNL; Spellman, Donald J [ORNL; Bevard, Bruce Balkcom [ORNL

2009-01-01T23:59:59.000Z

106

Gas-cooled fast breeder reactor steady-state irradiation testing program  

Science Conference Proceedings (OSTI)

The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast breeder reactor program are considered, and available irradiation facilities are examined.

Acharya, R.T.; Campana, R.J.; Langer, S.

1980-08-01T23:59:59.000Z

107

Tests of candidate materials for particle bed reactors  

DOE Green Energy (OSTI)

Rhenium metal hot frits and zirconium carbide-coated fuel particles appear suitable for use in flowing hydrogen to at least 2000 K, based on previous tests. Recent tests on alternate candidate cooled particle and frit materials are described. Silicon carbide-coated particles began to react with rhenium frit material at 1600 K, forming a molten silicide at 2000 K. Silicon carbide was extensively attacked by hydrogen at 2066 K for 30 minutes, losing 3.25% of its weight. Vitrous carbon was also rapidly attacked by hydrogen at 2123 K, losing 10% of its weight in two minutes. Long term material tests on candidate materials for closed cycle helium cooled particle bed fuel elements are also described. Surface imperfections were found on the surface of pyrocarbon-coated fuel particles after ninety days exposure to flowing (approx.500 ppM) impure helium at 1143 K. The imperfections were superficial and did not affect particle strength.

Horn, F.L.; Powell, J.R.; Wales, D.

1987-01-01T23:59:59.000Z

108

Results and Analyses of Irradiation/Anneal Experiments Conducted on Yankee Rowe Reactor Pressure Vessel Surrogate Materials: Yankee Atomic Electric Company Test Reactor Program  

Science Conference Proceedings (OSTI)

Many variables influence the response of reactor vessel steels to neutron irradiation. This study looks at the influence of irradiation temperature, steel heat treatment and microstructure, and nickel and phosphorus content on the irradiation response of high-copper reactor vessel steel. Also addressed are several studies evaluating the potential of thermal annealing to restore the mechanical properties of the steels tested.

1996-03-22T23:59:59.000Z

109

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report  

SciTech Connect

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

Lee, D.D.; Collins, J.L.

2000-02-01T23:59:59.000Z

110

A review of two recent occurrences at the Advanced Test Reactor involving subcontractor activities  

Science Conference Proceedings (OSTI)

This report documents the results of a brief, unofficial investigation into two incidents at the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR) facility, reported on October 25 and 31, 1997. The first event was an unanticipated breach of confinement. The second involved reactor operation with an inoperable seismic scram subsystem, violating the reactor`s Technical Specifications. These two incidents have been found to be unrelated. A third event that occurred on December 16, 1996, is also discussed because of its similarities to the first event listed above. Both of these incidents were unanticipated breaches of confinement, and both involved the work of construction subcontractor personnel. The cause for the subcontractor related occurrences is a work control process that fails to effectively interface with LMITCO management. ATR Construction Project managers work sufficient close with construction subcontractor personnel to understand planned day-to-day activities. They also have sufficient training and understanding of reactor operations to ensure adherence to applicable administrative requirements. However, they may not be sufficiently involved in the work authorization and control process to bridge an apparent communications gap between subcontractor employees and Facility Operations/functional support personnel for work inside the reactor facility. The cause for the inoperable seismic scram switch (resulting from a disconnected lead) is still under investigation. It does not appear to be subcontractor related.

Dahlke, H.J.; Jensen, N.C.; Vail, J.A.

1997-11-01T23:59:59.000Z

111

2012 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

Mike Lewis

2013-02-01T23:59:59.000Z

112

2011 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

Mike Lewis

2012-02-01T23:59:59.000Z

113

2010 Radiological Monitoring Results Associated with the Advance Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

mike lewis

2011-02-01T23:59:59.000Z

114

Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor  

SciTech Connect

This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

Lowry, N.J.

1998-10-21T23:59:59.000Z

115

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

116

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

117

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

118

Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor  

Science Conference Proceedings (OSTI)

Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

1984-06-01T23:59:59.000Z

119

Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine  

SciTech Connect

This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

Reilly, Raymond W.

2012-07-30T23:59:59.000Z

120

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors  

SciTech Connect

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

1993-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor  

SciTech Connect

A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

John Darrell Bess

2009-05-01T23:59:59.000Z

122

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

Science Conference Proceedings (OSTI)

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Gusev, S. I. [JSC 'FSK EES' (Russian Federation); Karpov, V. N.; Kiselev, A. N.; Kochkin, V. I. [Scientific-Research Institute of Electric Power Engineering (VNIIE) - Branch of the JSC 'NTTs Elektroenergetiki', (Russian Federation)

2009-09-15T23:59:59.000Z

123

Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications  

DOE Green Energy (OSTI)

Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

1981-10-01T23:59:59.000Z

124

Initial data testing of ENDF/B-VI for thermal reactor benchmark analysis  

SciTech Connect

This paper summarizes some early data testing of ENDF/B-VI by members of the Cross Section Evaluation Working Group (CSEWG) Thermal Reactor Data Testing Subcommittee. Projections of ENDF/B-VI performance in thermal benchmark calculations are beginning to be available; and in some cases the calculations were performed with only a portion of the cross sections taken from version VI, the remainder taken from earlier data files. A factor delaying the thermal reactor data testing is that the final {sup 235}U evaluation has not yet been officially released--only an earlier evaluation with a constant low-energy eta value (like in version V) is currently available. The official version VI {sup 235}U evaluation (scheduled for release as Mod-1) gives a drooping eta variation at low energy; i.e., eta decreases with decreasing energy. This behavior was suggested by European studies to improve the calculation of temperature coefficients in LWRs.

Williams, M.L. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Kahler, A.C. [Bettis Atomic Power Lab., West Mifflin, PA (United States); MacFarlane, R.E. [Los Alamos National Lab., NM (United States); Milgram, M. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Wright, R.Q. [Oak Ridge National Lab., TN (United States)

1991-12-31T23:59:59.000Z

125

Results of the DF-4 BWR (boiling water reactor) control blade-channel box test  

DOE Green Energy (OSTI)

The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.

Gauntt, R.O.; Gasser, R.D.

1990-10-01T23:59:59.000Z

126

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor  

SciTech Connect

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams.

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon [Massachusetts Institute of Technology (United States)

2005-05-15T23:59:59.000Z

127

An Experimental Shield Test Facility for the Development of Minimum Weight Shields for Compact Reactor Power Systems  

SciTech Connect

Discussions are given of the characteristics of fission-source plate, graphite reactor, and pool-type reactor facilities applicable to development studies of minimum weight shielding materials. Advantages of a proposed SNAP dual-purpose shielding facility are described in terms of a disk-shaped fission-source plate, reactor, and building. A program for the study of advanced shielding materials is discussed for materials and configuations to be evaluted with the fission-source plate, the testing of the prototype at high-power levels, and full-power tests on the actual reactor.

Tomlinson, R.L.

1959-08-07T23:59:59.000Z

128

Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements  

SciTech Connect

The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

Leland M. Montierth

2010-12-01T23:59:59.000Z

129

Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors  

SciTech Connect

Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

2012-02-01T23:59:59.000Z

130

Initial confinement studies of ohmically heated plasmas in the tokamak fusion test reactor  

DOE Green Energy (OSTI)

Initial operation of the tokamak fusion test reactor has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1--0.2 s for a line-average density range (n-bar/sub e/) of (1--2.5) x 10/sup 19/ m/sup -3/ with electron temperatures of T/sub e/(o)approx.1.2--2.2 keV, ion temperatures of T/sub i/(0)approx.0.9--1.5 keV, and Z/sub eff/approx.3. A comparison of Princeton large torus, poloidal divertor experiment, and tokamak fusion test reactor plasma confinement supports a dimension-cubed scaling law.

Efthimion, P.C.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.L.; Coonrod, J.; Davis, S.; Dylla, H.F.; Fonck, R.; Furth, H.P.

1984-04-23T23:59:59.000Z

131

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

132

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

SciTech Connect

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

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

2012-08-01T23:59:59.000Z

133

Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility  

Science Conference Proceedings (OSTI)

A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

2008-04-01T23:59:59.000Z

134

Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency  

SciTech Connect

This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project.

Hans Gougar

2010-02-01T23:59:59.000Z

135

Page not found | Department of Energy  

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

11 - 23120 of 28,905 results. 11 - 23120 of 28,905 results. Download CX-008347: Categorical Exclusion Determination West Hackberry Building 301 Air Handling Unit-1 Heating, Ventilation, and Air Conditioning Replacement CX(s) Applied: B1.4 Date: 04/16/2012 Location(s): Louisiana Offices(s): Strategic Petroleum Reserve Field Office http://energy.gov/nepa/downloads/cx-008347-categorical-exclusion-determination Article Recovery Act's HWCTR Project Empty of Equipment, Ready for Grouting AIKEN, S.C. - Thanks to investments from the American Recovery and Reinvestment Act, the next phase has begun on decommissioning the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site (SRS). http://energy.gov/em/articles/recovery-act-s-hwctr-project-empty-equipment-ready-grouting Article DOE Completes Disposal Operations In Panel 5 of the WIPP

136

NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS  

Science Conference Proceedings (OSTI)

This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

Hakan Ozaltun & Herman Shen

2011-11-01T23:59:59.000Z

137

CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors  

Science Conference Proceedings (OSTI)

An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

Carcreff, H. [Alternative Energies and Atomic Energy Commission CEA, Saclay Center, DEN/DANS/DRSN/SIREN, Gif Sur Yvette, 91191 (France)

2011-07-01T23:59:59.000Z

138

Run - Beyond - Cladding - Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program  

Science Conference Proceedings (OSTI)

In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab.

Batte, G.L. (Argonne National Lab., Idaho Falls, ID (USA)); Hoffman, G.L. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

139

Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

Donna P. Guillen

2012-07-01T23:59:59.000Z

140

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility  

SciTech Connect

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

2013-02-01T23:59:59.000Z

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141

Status of the NGNP Fuel Experiment AGR-2 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), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

Blaine Grover

2012-10-01T23:59:59.000Z

142

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology  

Science Conference Proceedings (OSTI)

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team projects and faculty/staff exchanges. In June of 2008, the first week-long ATR NSUF Summer Session was attended by 68 students, university faculty and industry representatives. The Summer Session featured presentations by 19 technical experts from across the country and covered topics including irradiation damage mechanisms, degradation of reactor materials, LWR and gas reactor fuels, and non-destructive evaluation. High impact research results from leveraging the entire research infrastructure, including universities, industry, small business, and the national laboratories. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. Current partner facilities include the MIT Reactor, the University of Michigan Irradiated Materials Testing Laboratory, the University of Wisconsin Characterization Laboratory, and the University of Nevada, Las Vegas transmission Electron Microscope User Facility. Needs for irradiation of material specimens at tightly controlled temperatures are being met by dedication of a large in-pile pressurized water loop facility for use by ATR NSUF users. Several environmental mechanical testing systems are under construction to determine crack growth rates and fracture toughness on irradiated test systems.

T. R. Allen; J. B. Benson; J. A. Foster; F. M. Marshall; M. K. Meyer; M. C. Thelen

2009-05-01T23:59:59.000Z

143

Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor  

SciTech Connect

A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL.

Zeuch, W.R.; A-Moneim, M.T.

1979-01-01T23:59:59.000Z

144

Bench-scale reactor tests of low-temperature, catalytic gasification of wet, industrial wastes  

DOE Green Energy (OSTI)

Bench-scale reactor tests are under way at Pacific Northwest Laboratory to develop a low-temperature, catalytic gasification system. The system, licensed under the trade name Thermochemical Environmental Energy System (TEES{reg sign}), is designed for to a wide variety of feedstocks ranging from dilute organics in water to waste sludges from food processing. The current research program is focused on the use of a continuous-feed, tubular reactor. The catalyst is nickel metal on an inert support. Typical results show that feedstocks such as solutions of 2% para-cresol or 5% and 10% lactose in water or cheese whey can be processed to >99% reduction of chemical oxygen demand (COD) at a rate of up to 2 L/hr. The estimated residence time is less than 5 min at 360{degree}C and 3000 psig, not including 1 to 2 min required in the preheating zone of the reactor. The liquid hourly space velocity has been varied from 1.8 to 2.9 L feedstock/L catalyst/hr depending on the feedstock. The product fuel gas contains 40% to 55% methane, 35% to 50% carbon dioxide, and 5% to 10% hydrogen with as much as 2% ethane, but less than 0.1% ethylene or carbon monoxide, and small amounts of higher hydrocarbons. The byproduct water stream carries residual organics amounting to less than 500 mg/L COD. 9 refs., 1 fig., 4 tabs.

Elliott, D.C.; Neuenschwander, G.G.; Baker, E.G.; Butner, R.S.; Sealock, L.J.

1990-04-01T23:59:59.000Z

145

Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop  

Science Conference Proceedings (OSTI)

This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

Donna Post Guillen

2012-11-01T23:59:59.000Z

146

IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR  

SciTech Connect

The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.

David W. Nigg; Joseph W. Nielsen; Benjamin M. Chase; Ronnie K. Murray; Kevin A. Steuhm

2012-04-01T23:59:59.000Z

147

Improved computational neutronics methods and validation protocols for the advanced test reactor  

SciTech Connect

The Idaho National Laboratory (INL) is in the process of updating the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purposes. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry have been conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for flexible and repeatable ATR physics code validation protocols that are consistent with applicable national standards. (authors)

Nigg, D. W.; Nielsen, J. W.; Chase, B. M.; Murray, R. K.; Steuhm, K. A.; Unruh, T. [Idaho National Laboratory, 2525 Fremont Street, Idaho Falls, ID 83415-3870 (United States)

2012-07-01T23:59:59.000Z

148

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

149

Measurements of Nonlinear Energy Transfer in Turbulence in the Tokamak Fusion Test Reactor  

SciTech Connect

The application of a new bispectral analysis technique to density fluctuation measurements in the core of the Tokamak Fusion Test Reactor indicates that the peak in the autopower spectrum usually lies in a region of linear stability. Large changes in the linear and nonlinear characteristics of the turbulence are observed as the plasma toroidal rotation and/or confinement properties are varied, while estimates of the turbulence-driven diffusivity varies only slightly with rotation. These observations are consistent with the operation of a global organizing property that may be related to the observation of Bohm-like scaling of ion thermal transport. {copyright} {ital 1997} {ital The American Physical Society}

Kim, J.S.; Fonck, R.J.; Durst, R.D. [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States)] [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States); Fernandez, E.; Terry, P.W. [Department of Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States)] [Department of Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States); Paul, S.F.; Zarnstorff, M.C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)] [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

1997-08-01T23:59:59.000Z

150

In-situ Creep Testing Capability Development for Advanced Test Reactor  

SciTech Connect

Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2010-08-01T23:59:59.000Z

151

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

152

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

153

Hot-gas filter testing with the transport reactor demonstration unit  

Science Conference Proceedings (OSTI)

The objectives of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Energy & Environmental Research Center (EERC) is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot-gas filter element performance (particulate collection efficiency, filter pressure differential, filter cleanability, and durability) as a function of temperature and filter face velocity during short-term operation (100-200 hours). This filter vessel will be utilized in combination with the TRDU to evaluate the performance of selected hot-gas filter elements under gasification operating conditions. This work will directly support the power systems development facility (PSDF) utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and, indirectly, the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville.

Mann, M.D.; Swanson, M.L.; Ness, R.O.; Haley, J.S.

1995-11-01T23:59:59.000Z

154

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations  

Science Conference Proceedings (OSTI)

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

J. L. Rempe; D. L. Knudson; J. E. Daw

2011-03-01T23:59:59.000Z

155

Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor  

SciTech Connect

Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

1983-01-01T23:59:59.000Z

156

Hot-Gas Filter Testing with a Transport Reactor Development Unit  

Science Conference Proceedings (OSTI)

The objective of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Environmental Research Center is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot- gas filter elements as a function of particulate collection efficiency, filter pressure differential, filter cleanability, and durability during relatively short-term operation (100-200 hours). A filter vessel will be used in combination with the TRDU to evaluate the performance of selected hot- gas filter elements under gasification operating conditions. This work will directly support the Power Systems Development Facility utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and indirectly the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville and the Clean Coal IV Pinon Pine IGCC Power Project. This program has a phased approach involving modification and upgrades to the TRDU and the fabrication, assembly, and operation of a hot-gas filter vessel (HGFV) capable of operating at the outlet design conditions of the TRDU. Phase 1 upgraded the TRDU based upon past operating experiences. Additions included a nitrogen supply system upgrade, upgraded LASH auger and 1807 coal feed lines, the addition of a second pressurized coal feed hopper and a dipleg ash hopper, and modifications to spoil the performance of the primary cyclone. Phase 2 included the HGFV design, procurement, and installation. Phases 3 through 5 consist of 200-hour hot-gas filter tests under gasification conditions using the TRDU at temperatures of 540-650{degrees}C (1000-1200{degrees}F), 9.3 bar, and face velocities of 1.4, 2. and 3.8 cm/s, respectively. The increased face velocities are achieved by removing candles between each test.

Swanson, M.L.; Ness, R.O., Jr. [North Dakota Univ., Grand Forks, ND (United States). Energy and Environmental Research Center

1996-12-31T23:59:59.000Z

157

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

158

Production test IP-278-A: Verification of BPA loss bulk temperature surge at the DE-Reactor. Supplement A  

SciTech Connect

This report details planning to run a second outage test at the DR-Reactor using the same instrumentation and procedure as an earlier test but increasing the trip-out level from 800 MW up to a maximum of 1200 MW.

Jones, S.S.

1960-01-07T23:59:59.000Z

159

Interim report VII, production test IP-549-A half-plant low alum feed water treatment at F Reactor  

SciTech Connect

A half-plant low alum water treatment test began at F Reactor on January 16, 1963. The test, which had been prompted by the analysis of ledge corrosion attack on fuel elements, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge and groove corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from the normal production fuel irradiated in process water treated with two different alum feed rates. Six 20-column fuel discharges, ten columns from each side of the reactor, have been taken during the test as follows: (1) One discharge prior to the start of the test. (2) One discharge such that the test side was exposed to coolant treated with both high and low alum feed. (3) Four discharges under test conditions. This report discusses the results obtained from the fifth discharge under test conditions.

Geier, R.G.

1964-03-18T23:59:59.000Z

160

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010  

SciTech Connect

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. 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).

Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

2010-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" from the National Library of EnergyBeta (NLEBeta).
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161

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013  

SciTech Connect

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. 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).

David W. Nigg

2013-09-01T23:59:59.000Z

162

Engineering considerations in the selection of the tokamak to follow the Tokamak Fusion Test Reactor (TFTR)  

SciTech Connect

The tokamak to follow the Tokamak Fusion Test Reactor (TFTR) should satisfy two important objectives. First, it should be a significant step in physics and engineering goals in order to maintain the level of progress which the US has established as the world leader in fusion energy development. The second objective should be to provide the information necessary to support the strategy and goals of the long-range Department of Energy (DOE) Fusion Program. In their Comprehensive Program Management Plan, the DOE identifies the need for a reactor technology program in the 1990s in which the major goal is to prove engineering feasibility. In this paper, the specific engineering needs are identified which have been developed through the tokamak design studies over the past decade. On the basis of these needs, it appears that several options are available for the next tokamak to follow TFTR. The final choice of the concept will involve consideration of the technical needs and the reality of the Fusion Program budget.

Shannon, T.E.

1983-01-01T23:59:59.000Z

163

Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*  

Science Conference Proceedings (OSTI)

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja

1994-05-01T23:59:59.000Z

164

THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS  

DOE Green Energy (OSTI)

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

David S. Duncan; Vondell J. Balls; Stephanie L. Austad

2008-09-01T23:59:59.000Z

165

Initial confinement studies of ohmically heated plasmas in the Tokamak Fusion Test Reactor  

DOE Green Energy (OSTI)

Initial operation of the Tokamak Fusion Test Reactor (TFTR) has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1 to 0.2 s for a line-average density range (anti n/sub e/) of 1 to 2.5 x 10/sup 19/ m/sup -3/ with electron temperatures of T/sub e/(o) approx. 1.2 to 2.2 keV, ion temperatures of T/sub i/(o) approx. 0.9 to 1.5 keV, and Z/sub eff/ approx. 3. A comparison of PLT, PDX, and TFTR plasma confinement supports a dimension-cubed scaling law.

Efthimion, P.C.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.L.; Coonrod, J.; Davis, S.; Dylla, H.F.; Fonck, R.; Furth, H.P.

1984-06-01T23:59:59.000Z

166

Short Term Irradiation Test of Fuel Containing Minor Actinides Using the Experimental Fast Reactor Joyo  

Science Conference Proceedings (OSTI)

A mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast rector Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted as part of the short-term phase of this program in May and August 2006. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), and MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX). The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes. After 10 minutes irradiation test, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins with neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. The linear heat rate for each MA-MOX test fuel pin was calculated using the Monte Carlo calculation code MCNP. The calculated fission rates were compared with the measured data based on the Nd-148 method. The maximum linear heat rate was approximately 444{+-}19 W/cm at the actual reactor power of 119.6 MWt. Post irradiation examination of these pins to confirm the absence of fuel melting and the local concentration under irradiation of NpO{sub 2-x} or AmO{sub 2-x}, in the (U,Pu)0{sub 2-x}, fuel are underway. The test results are expected to reduce uncertainties on the margin in the thermal design for MA-MOX fuel. (authors)

Sekine, Takashi; Soga, Tomonori; Koyama, Shin-ichi; Aoyama, Takafumi [Oarai Research and Development Center, Japan Atomic Energy Agency. 4002 Narita, Oarai, Ibaraki 311-1393 (Japan); Wootan, David [Pacific Northwest National Laboratoy, M/S K8-34, P.O. Box 999 Richland, WA 99352 (United States)

2007-07-01T23:59:59.000Z

167

Advanced Test Reactor LEU Fuel Conversion Feasibility Study -- 2006 Annual Report  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the U.S. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U 235 enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U 235 loading in the LEU core, such that the differences in Keff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The Monte-Carlo coupled with ORIGEN2 (MCWO) depletion methodology was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the Keff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (OSCC, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

G. S. Chang; R. G. Ambrosek

2006-10-01T23:59:59.000Z

168

Advanced Test Reactor LEU Fuel Conversion Feasibility Study (2006 Annual Report)  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The depletion methodology, Monte-Carlo coupled with ORIGEN2 (MCWO), was used to calculate K-eff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders (OSCCs), safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

Gray S. Chang; Richard G. Ambrosek; Misti A. Lillo

2006-12-01T23:59:59.000Z

169

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. INITIAL FULL POWER AND LIMITED ENDURANCE TESTS OF THE ML-1 NUCLEAR POWER PLANT. Final Test Report  

SciTech Connect

The evaluation of the data generated during the full power and limited endurance tests of the ML-1 mobile nuclear power plant indicates that the reactor performs in accordance with the design specifications. During the 101 hr test period, the reactor attained a maximum power of 3.44 Mw( and 247 kw(e) was measured at the output shaft of the turbine-compressor set. No operating limits were exceeded during these tests and all systems performed satisfactorily Except for the known performance deficiency of the turbinecompressor set, which prevented the attainment of design output power, no operational, stability, or control problems were encountered. All test objectives were achieved and the tests were considered completely successful. (auth)

Kattchee, N.

1963-07-01T23:59:59.000Z

170

Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions  

DOE Green Energy (OSTI)

The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

2007-03-30T23:59:59.000Z

171

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

172

Testing of a 7-tube palladium membrane reactor for potential use in TEP  

SciTech Connect

A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO{sub 2} ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m{sup 2} and a catalyst volume to membrane area ratio of 4.63 cc/cm{sup 2} (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m{sup 2}). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m{sup 2}. The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm{sup 2}. The total membrane area of the 7-tube PMR (0.0851 m{sup 2}) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m{sup 2}). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR scaling method; (5) Evaluate PMR with realistic feed compositions; (6) Evaluate PMR performance with varying permeate pressures; (7) Study coking-related issues; and (8) Identify any unexpected behavior that may require further investigation (used to study transient behavior). This report presents the tests results defined by these objectives.

Carlson, Bryan J [Los Alamos National Laboratory; Trujillo, Stephen [Los Alamos National Laboratory; Willms, R. Scott [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

173

TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR  

SciTech Connect

As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten-rhenium and platinum rhodium thermocouples can be avoided. INL is also developing an Ultrasonic Thermometry (UT) capability. In addition to small size, UT’s offer several potential advantages over other temperature sensors. Measurements may be made near the melting point of the sensor material, potentially allowing monitoring of temperatures up to 3000 C. In addition, because no electrical insulation is required, shunting effects are avoided. Most attractive, however, is the ability to introduce acoustic discontinuities to the sensor, as this enables temperature measurements at several points along the sensor length. As discussed in this paper, the suite of temperature monitors offered by INL is not only available to ATR users, but also to users at other MTRs.

J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

2012-03-01T23:59:59.000Z

174

Deviation to the Test Program and Procedures for the 710 Critical Experiment Reactor Control Drum Mockup Experiment  

SciTech Connect

This document describes a deviation from the "Test Program and Procedures for the 710 Critical Experiment Reactor Control Drum Mockup Experiment," TM-64-3-706, which was made in accordance with ITS Standard Practice J80-81 on September 14, 1964. The deviation did not involve a significant change in the safety of the operation.

Sims, F.L.

1964-09-14T23:59:59.000Z

175

Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington  

SciTech Connect

During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensee’s final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

S.J. Roberts

2007-03-20T23:59:59.000Z

176

System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)  

SciTech Connect

This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

Jeffrey Bryan

2009-06-01T23:59:59.000Z

177

Monitoring and Control Research Using a University Reactor and SBWR Test-Loop  

Science Conference Proceedings (OSTI)

The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

Robert M. Edwards

2003-09-28T23:59:59.000Z

178

Thermal analysis for a spent reactor fuel storage test in granite  

Science Conference Proceedings (OSTI)

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

Montan, D.N.

1980-09-01T23:59:59.000Z

179

Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor  

SciTech Connect

The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

2012-02-01T23:59:59.000Z

180

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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181

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

182

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

183

Investigation of global Alfven instabilities in the Tokamak Fusion Test Reactor  

SciTech Connect

Toroidal Alfven eigenmodes (TAE) were excited by the energetic neutral beam ions tangentially injected into plasmas at low magnetic field in the Tokamak Fusion Test Reactor (TFTR) ({ital Proceedings} {ital of} {ital the} 11{ital th} {ital International} {ital Conference} {ital on} {ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Fusion} {ital Research} (IAEA, Vienna, 1987), Vol. 1, p. 51). The injection velocities were comparable to the Alfven speed. The modes were identified by measurements from Mirnov coils and beam emission spectroscopy (BES). TAE modes appear in bursts whose repetition rate increases with beam power. The neutron emission rate exhibits sawtoothlike behavior and the crashes always coincide with TAE bursts. This indicates ejection of fast ions from the plasma until these modes are stabilized. The dynamics of growth and stabilization were investigated at various plasma currents and magnetic fields. The results indicate that the instability can effectively clamp the number of energetic ions in the plasmas. The observed instability threshold is discussed in light of recent theories. In addition to these TAE modes, intermittent oscillations at three times the fundamental TAE frequency were observed by Mirnov coils, but no corresponding signal was found in BES. It appears that these high-frequency oscillations do not have a direct effect on the plasma neutron source strength.

Wong, K.L.; Durst, R.; Fonck, R.J.; Paul, S.F.; Roberts, D.R.; Fredrickson, E.D.; Nazikian, R.; Park, H.K.; Bell, M.; Bretz, N.L.; Budny, R.; Cheng, C.Z.; Cohen, S.; Hammett, G.W.; Jobes, F.C.; Johnson, L.; Meade, D.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Sabbagh, S.; Synakowski, E.J. (Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States))

1992-07-01T23:59:59.000Z

184

Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program  

Science Conference Proceedings (OSTI)

This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II.

Batte, G.L.; Pahl, R.G. [Argonne National Lab., Idaho Falls, ID (United States); Hofman, G.L. [Argonne National Lab., IL (United States)

1993-09-01T23:59:59.000Z

185

EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR  

Science Conference Proceedings (OSTI)

The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

2011-03-01T23:59:59.000Z

186

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011  

SciTech Connect

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. 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). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.

David W. Nigg; Devin A. Steuhm

2011-09-01T23:59:59.000Z

187

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011  

SciTech Connect

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. 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). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.

David W. Nigg; Devin A. Steuhm

2011-09-01T23:59:59.000Z

188

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012  

SciTech Connect

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. 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). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.

David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

2012-09-01T23:59:59.000Z

189

Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system  

Science Conference Proceedings (OSTI)

The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

Dautel, W.A.

1996-10-01T23:59:59.000Z

190

Materials Reliability Program, Reactor Vessel Head Boric Acid Corrosion Testing (MRP-165)  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential...

2005-12-14T23:59:59.000Z

191

Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF  

SciTech Connect

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could actually help extend the life of a fuel pin in a “long pin” reactor design to a higher peak fuel burnup.

D. L. Porter; H.C. Tsai

2012-08-01T23:59:59.000Z

192

Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)  

E-Print Network (OSTI)

1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

American Society for Testing and Materials. Philadelphia

2003-01-01T23:59:59.000Z

193

Solar reforming of methane in a direct absorption catalytic reactor on a parabolic dish: I-test and analysis  

DOE Green Energy (OSTI)

The concept of solar driven chemical reaction in a commercial-scale volumetric receiver/reactor on a parabolic concentrator was successfully demonstrated in the CAtalytically Enhanced Solar Absorption Receiver (CAESAR) test. Solar reforming of methane (CH[sub 4]) with carbon dioxide (CO[sub 2]) was achieved in a 64 cm diameter direct absorption reactor on a parabolic dish capable of 150 kW solar power. The reactor was a catalytic volumetric absorber consisting of a multilayered, porous alumina foam disk coated with rhodium (Rh) catalyst. The system was operated during both steady-state and solar transient (cloud passage) conditions. The total solar power absorbed reached values up to 97 kW and the maximum methane conversion was 70%. Receiver thermal efficiencies ranged up to 85% and chemical efficiencies peaked at 54%. The absorber performed satisfactorily in promoting the reforming reaction during the tests without carbon formation. However, problems of cracking and degradation of the porous matrix, nonuniform dispersion of the Rh through the absorber, the catalyst deactivation due to sintering and possible encapsulation, must be resolved to achieve long-term operation and eventual commercialization.

Muir, J.F.; Hogan, R.E. Jr.; Skocypec, R.D. (Sandia National Lab., Albuquerque, NM (United States)); Buck, R. (DLR-ITT, Stuttgart (Germany))

1994-06-01T23:59:59.000Z

194

Solar test of an integrated sodium reflux heat-pipe receiver/reactor for thermochemical energy transport  

DOE Green Energy (OSTI)

In October 1987, a chemical reactor integrated into a sodium reflux heat-pipe receiver was tested in the solar furnace at the Weizmann Institute of Science, Rehovot, Israel. The reaction carried out was the carbon dioxide reforming of methane. This reaction is one of the leading candidates for thermochemical energy transport either within a distributed solar receiver system or over long distances. The Schaeffer Solar Furnace consists of a 96 square meter heliostat and a 7.3 meter diameter dish concentrator with a 65-degree rim angle and a 3.5 meter focal length. Measurements have shown a peak concentration ratio of over 10,000 and a total power of 15 kW at an insolation of 800 w/square meter. The receiver/reactor contains seven catalyst-filled tubes inside an evacuated metal box containing sodium. The front surface of this box serves as the solar absorber of the receiver. In operation, concentrated sunlight heats the 1/8-inch Inconel plate and vaporizes sodium from the wire-mesh wick attached to the back of it. The sodium vapor condenses on the reactor tubes, releases its latent heat, and returns by gravity to the wick. Test results and areas for future development are discussed.

Diver, R.B.; Fish, J.D.; Levitan, R.; Levy, M.; Rosin, H.; Richardson, J.T.

1988-01-01T23:59:59.000Z

195

How accurately can one test CPT conservation with reactor and solar neutrino experiments?  

E-Print Network (OSTI)

We show that the combined data from solar neutrino experiments and from the KamLAND reactor neutrino experiment can establish an upper limit on, or detect, potential CPT violation in the neutrino sector of order 10^{-20} GeV to 10^{-21} GeV.

John N. Bahcall; V. Barger; Danny Marfatia

2002-01-23T23:59:59.000Z

196

AGR-2: The first irradiation of French HTR fuel in Advanced Test Reactor  

SciTech Connect

AGR-2, the second irradiation of the US program for qualification of the NGNP fuel, is open to international participation within the scope of the Generation IV International Forum. In this frame, it includes in its multi-capsule irradiation rig an irradiation of French HTR fuel manufactured in the CAPRI line (GAIA facility at CEA/Cadarache and AREVA/CERCA compacting line at Romans). The AGR-2 irradiation is designed to place our first fabrications of HTR particles under operating conditions that are representative of ANTARES project while keeping close to the test range of the German fuel as much as possible, which is the reference in terms of irradiation behavior. A few batches of particles and 12 fuel compacts were produced and characterized in 2009 by CEA and CERCA. The fuel main characteristics are in conformity with our specifications and in compliance with INL requirements. The AGR-2 experiment is based on the design and devices used in the first experiment of the AGR program. The design makes it possible to monitor the irradiation conditions and in particular, the temperature, the power and the fission products released from fuel particles. The in pile equipment consists of a multi-capsule device designed to simultaneously irradiate six independent capsules with temperature control. The out-of-core part consists of the equipment for actively controlling temperature and measuring the fission products release on-line. The target conditions for the irradiation experiment were defined with the aim of comparing the results obtained under irradiation with German particles along with the objectives of reaching burn-up and fluence targets to validate the behavior of our fuel in a significant range (15% FIMA – 5 × 1025 n/m2 at 600 EFPD with centerline fuel temperature about 1100 degrees C). These conditions have to be representative of ANTARES project characteristics. These target conditions were compared with final results from neutron and thermal design studies performed by INL team, and preliminary thermal mechanical ATLAS calculations were carried out by CEA from this pre-design. Despite the mean burn-up achieved in approximately 600 EFPD being a little high (16.3% FIMA max. associated with a low fluence up to 2.85 × 1025 n/m2), this irradiation will nevertheless encompass the range of irradiation effects covered in our experimental objectives (maximum stress peak at start of irradiation then sign inversion of the stress in the SiC layer). In addition, the fluence and burn-up acceleration factors are very similar to those of the German reference experiments. This experimental irradiation began in July 2010 in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and first results have been acquired.

T. Lambert; B. Grover; P. Guillermier; D. Moulinier; F. Imbault Huart

2012-10-01T23:59:59.000Z

197

PROCEEDINGS OF THE AEC SYMPOSIUM FOR CHEMICAL PROCESSING OF IRRADIATED FUELS FROM POWER, TEST, AND RESEARCH REACTORS, RICHLAND, WASHINGTON, OCTOBER 20 AND 21, 1959  

SciTech Connect

A review is presented in this symposium of the technology currently available for processing spent fuels from research, test, and power reactors. Twenty-one papers are included. Separate abstracts have been prepared for each paper. (W.L.H.)

1960-01-01T23:59:59.000Z

198

10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement  

SciTech Connect

The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

David E. Korns

2012-05-01T23:59:59.000Z

199

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility  

Science Conference Proceedings (OSTI)

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

Not Available

1992-07-01T23:59:59.000Z

200

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

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201

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.  

Science Conference Proceedings (OSTI)

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30T23:59:59.000Z

202

FIRST SODIUM REACTOR EXPERIMENT (SRE) TEST OF HALLAM NUCLEAR POWER FACILITY (HNPF) CONTROL MATERIALS  

SciTech Connect

An experiment was conducted in the SRE to measure temperatures and neutron flux levels in and near a boron-containing simulated control rod. The data are being used to check analytical methods developed for prediction of control rod heat generation rates and maximum temperatures in this type of control rod in the Hallam Nuclear Power Facility. The maximum observed temperatures with a reactor power level of 20 Mw were 1363 deg F for a boron-- nickel alloy ring having a 0.105-in. radial clearance with the thimble and 1100 deg F for a boron -nickel alloy ring having a 0.020-in. radial clearance. The maximum temperature difference between the coolant and the control rod was 473 deg F. It is concluded that the expected greater heat generation rates in the Hallam reactor would prohibit the use of boron-containing absorber materials in a combined a him-safety rod. (auth)

Arneson, S.O.

1959-06-01T23:59:59.000Z

203

Variation of the magnetic susceptibility of artificial graphite with exposure in the materials testing reactor  

SciTech Connect

The magnetic susceptibility of artificial graphite was determined as a function of exposure in the MTR. Specimens were studied with exposures ranging from 0.07 to 82 {times} 10{sup18} nvt. Fluxes were determined by means of x-ray measurements and resistivity measurements. The dependence of the magnetic susceptibility on exposure in the MTR and also in a Hanford reactor are graphed, and an equivalence factor is calculated.

McCelland, J.D.

1955-02-23T23:59:59.000Z

204

Materials Reliability Program: Reactor Vessel Head Boric Acid Corrosion Testing (MRP-199)  

Science Conference Proceedings (OSTI)

PWR coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential stages of boric acid corros...

2007-06-27T23:59:59.000Z

205

On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor  

Science Conference Proceedings (OSTI)

IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

Ayman I. Hawari; Mohamed A. Bourham

2010-04-22T23:59:59.000Z

206

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

207

Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). 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, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. 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 sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

Blaine Grover

2012-10-01T23:59:59.000Z

208

Brief summary of reactor core component welding for the Fast Flux Test Facility (FFTF)  

SciTech Connect

Included are descriptions of welding methods and joint design, welding equipment, and qualification tests. (DG)

Brown, W.F.

1974-04-15T23:59:59.000Z

209

Thermal Hydraulic Analysis of a Reduced Scale High Temperature Gas-Cooled Reactor Test Facility and its Prototype with MELCOR  

E-Print Network (OSTI)

Pursuant to the energy policy act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the Very High Temperature Reactor (VHTR) that will become the Next Generation Nuclear Plant (NGNP). Although plans to build a demonstration plant at Idaho National Laboratories (INL) are currently on hold, a cooperative agreement on HTGR research between the U.S. Nuclear Regulatory Commission (NRC) and several academic investigators remains in place. One component of this agreement relates to validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform HTGR licensing analyses. Because the NRC has used MELCOR for LWR licensing in the past and because MELCOR was recently updated to include gas-cooled reactor physics models, MELCOR is among the system codes of interest in the cooperative agreement. The impetus for this thesis was a code-to-experiment validation study wherein MELCOR computer code predictions were to be benchmarked against experimental data from a reduced-scale HTGR testing apparatus called the High Temperature Test Facility (HTTF). For various reasons, HTTF data is not yet available from facility designers at Oregon State University, and hence the scope of this thesis was narrowed to include only computational studies of the HTTF and its prototype, General Atomics’ Modular High Temperature Gas-Cooled Reactor (MHTGR). Using the most complete literature references available for MHTGR design and using preliminary design information on the HTTF, MELCOR input decks for both systems were developed. Normal and off-normal system operating conditions were modeled via implementation of appropriate boundary and inititial conditions. MELCOR Predictions of system response for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) conditions were checked against nominal design parameters, physical intuition, and some computational results available from previous RELAP5-3D analyses at INL. All MELCOR input decks were successfully built and all scenarios were successfully modeled under certain assumptions. Given that the HTTF input deck is preliminary and was based on dated references, the results were altogether imperfect but encouraging since no indications of as yet unknown deficiencies in MELCOR modeling capability were observed. Researchers at TAMU are in a good position to revise the MELCOR models upon receipt of new information and to move forward with MELCOR-to-HTTF benchmarking when and if test data becomes available.

Beeny, Bradley 1988-

2012-12-01T23:59:59.000Z

210

A disposition strategy for highly enriched, aluminum-based fuel from research and test reactors  

SciTech Connect

The strategy proposed in this paper offers the Department of Energy an approach for disposing of aluminum-based, highly enriched uranium (HEU) spent fuels from foreign and domestic research reactors. The proposal is technically, socially, and economically sound. If implemented, it would advance US non-proliferation goals while also disposing of the spent fuel`s waste by timely and proven methods using existing technologies and facilities at SRS without prolonged and controversial storage of the spent fuel. The fuel would be processed through 221-H. The radioactive fission products (waste) would be treated along with existing SRS high level waste by vitrifying it as borosilicate glass in the Defense Waste Processing Facility (DWPF) for disposal in the national geological repository. The HEU would be isotopically diluted, during processing, to low-enriched uranium (LEU) which can not be used to make weapons, thus eliminating proliferation concerns. The LEU can be sold to fabricators of either research reactor fuel or commercial power fuel. This proposed processing-LEU recycle approach has several important advantages over other alternatives, including: Lowest capital investment; lowest net total cost; quickest route to acceptable waste form and final geologic disposal; and likely lowest safety, health, and environmental impacts.

McKibben, J.M.; Gould, T.H.; McDonell, W.R.; Bickford, W.E.

1994-11-01T23:59:59.000Z

211

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters  

Science Conference Proceedings (OSTI)

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Konzek, G.J.

1983-07-01T23:59:59.000Z

212

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.  

SciTech Connect

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodate an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the

Feldman, E. (Nuclear Engineering Division)

2011-06-09T23:59:59.000Z

213

Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.  

Science Conference Proceedings (OSTI)

The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

2011-06-07T23:59:59.000Z

214

Direct Test of the Time-Independence of Fundamental Nuclear Constants Using the Oklo Natural Reactor  

E-Print Network (OSTI)

[NOTE: This 1983 preprint is being uploaded to arXiv.org after the death of its author, who supported online distribution of his work. Contact info of the submitter is at http://ilya.cc .] The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

Alexander I. Shlyakhter

2003-07-03T23:59:59.000Z

215

Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.  

DOE Green Energy (OSTI)

As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong 3-D effects result in large heat flux, temperature, and heat transfer variations around the tube wall; (b) there is a large difference in the heat transfer coefficient predicted by turbulence models and heat transfer correlations, and this underscores the need of experimental work to validate the thermal performance of the RCCS; and (c) tests at the NSTF would embody all important fluid flow and heat transfer phenomena in the RCCS, in addition to covering the entire parameter ranges that characterize these phenomena. Additional supporting scaling study results are available in Reference 2. The purpose of this work is to develop a high-level engineering plan for mechanical and instrumentation modifications to NSTF in order to meet the following two technical objectives: (1) provide CFD and system-level code development and validation data for the RCCS under prototypic (full-scale) natural convection flow conditions, and (2) support RCCS design validation and optimization. As background for this work, the report begins by providing a summary of the original NSTF design and operational capabilities. Since the facility has not been actively utilized since the early 1990's, the next step is to assess the current facility status. With this background material in place, the data needs and requirements for the facility are then defined on the basis of supporting analysis activities. With the requirements for the facility established, appropriate mechanical and instrumentation modifications to NSTF are then developed in order to meet the overall project objectives. A cost and schedule for modifying the facility to satisfy the RCCS data needs is then provided.

Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

2005-09-01T23:59:59.000Z

216

Review of deuterium--tritium results from the Tokamak Fusion Test Reactor  

SciTech Connect

The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-{beta}{sub {ital p}} discharge following a current rampdown. The fusion power density in a core of the plasma is {approx}2.8 MW m{sup {minus}3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) at 1500 MW total fusion power. The energy confinement time, {tau}{sub {ital E}}, is observed to increase in D--T, relative to D plasmas, by 20% and the {ital n}{sub {ital i}}(0) {ital T}{sub {ital i}}(0) {tau}{sub {ital E}} product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-{beta}{sub {ital p}} discharges. Ion cyclotron range of frequencies (ICRF) heating of a D--T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D--T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

McGuire, K.M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, C.W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M.; Bell, M.G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C.S.; Cheng, C.Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P.C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Fujita, T.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Hawryluk, R.J.; Heidbrink, W.; Herrmann, H.W.; Hill, K.W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Loughlin, M.J.; Machuzak, J.; Majeski, R.; Mansfield, D.K.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M.P.; McChesney, J.; McCormack, B.; McCune, D.C.; McKee, G.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mirnov, S.V.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Norris, M.; O`Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Parks, P.; Paul, S.F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A.L.; Ruskov, E.; Sabbagh, S.A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G.

1995-06-01T23:59:59.000Z

217

First-wall, blanket, and shield engineering test program for magnetically confined fusion power reactors  

Science Conference Proceedings (OSTI)

The key engineering areas identified for early study relate to FW/B/S system thermal-hydraulics, thermomechnics, nucleonics, electromagnetics, assembly, maintenance, and repair. Programmatic guidance derived frm planning exercises involving over thirty organizations (laboratories, industries, and universities) has indicated (1) that meaningful near term engineering testing should be feasible within the bounds of a modest funding base, (2) that there are existing facilities and expertise which can be profitably utilized in this testing, and (3) that near term efforts should focus on the measurement of engineering data and the verification/calibration of predictive methods for anticipated normal operational and transient FW/B/S conditions. The remainder of this paper discusses in more detail the planning strategies, proposed approach to near term testing, and longer range needs for integrated FW/B/S test facilities.

Maroni, V.A.

1980-01-01T23:59:59.000Z

218

Power Ramp Testing of Additive Fuel Rods in the Halden Reactor  

Science Conference Proceedings (OSTI)

Additive fuel provides an effective remedy against pellet-cladding interaction (PCI) failure and a means of mitigating severe secondary degradation in the form of long axial splits in the cladding after fuel failure. This report—part of EPRI's Fuel Reliability Program—summarizes five years worth of data on additive fuel properties, pre-irradiation corrosion and creep tests, microstructure characterization, and subsequent in-core power ramp tests performed on segments retrieved after two cycle...

2003-11-14T23:59:59.000Z

219

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core  

Science Conference Proceedings (OSTI)

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO/sub 2/ and UO/sub 2//metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO/sub 2/ crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process.

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01T23:59:59.000Z

220

Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables  

SciTech Connect

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

Lisa Harvego; Brion Bennett

2011-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

New Sensors for In-Pile Temperature Measurement at the Advanced Test Reactor National Scientific User Facility  

SciTech Connect

The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) in April 2007 to support U.S. research in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

J. L. Rempe; D. L. Knudson; J. E. Daw; K. G. Condie

2011-09-01T23:59:59.000Z

222

03/01/2006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http://english.people.com.cn/200603/01/print20060301_247035.html  

E-Print Network (OSTI)

the commercialized nuclear reactors in the world were designed for fission, a process contrary to the ITER's fusion03/01/2006 09:51 AMLoading "People's Daily Online -- Chinese experimental thermonuclear reactor experimental thermonuclear reactor on discharge test in July China's new generation experimental Tokamak fusion

223

Aspects of design, test and validation of the software for a computerized reactor protection system  

Science Conference Proceedings (OSTI)

In safety-oriented applications, the software has to fulfil certain stringent reliability requirements. In order to determine the reliability of the software, a variety of different methods can be used. The methods used for the reliability proof of a ... Keywords: Program testing, Proof of correctness, Protection system

U. Voges

1976-10-01T23:59:59.000Z

224

Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

Timothy A. Hyde

2012-06-01T23:59:59.000Z

225

ADMINISTRATION OF ORNL RESEARCH REACTORS  

SciTech Connect

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Casto, W.R.

1962-08-20T23:59:59.000Z

226

A tritium-compatible piezoelectric valve for the tokamak fusion test reactor  

SciTech Connect

This work describes modifications made to a commercial piezoelectric valve to make it sufficiently tritium compatible for the TFTR trritium injection scenario. The results of testing the valve for leakage and performance following a series of progressively more severe tritium exposures are also presented. Finally, a proposal for a totally radiation-compatible piezoelectric valve, suitable for tritium-burning fusion machines of the future, is decribed. 9 refs., 2 figs., 1 tab.

Coffin, D.O.; Cole, S.P.; Wilhelm, R.C.

1988-02-01T23:59:59.000Z

227

EM Recovery Act Press Releases | Department of Energy  

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

February 23, 2011 February 23, 2011 Recovery Act's HWCTR Project Empty of Equipment, Ready for Grouting AIKEN, S.C. - Thanks to investments from the American Recovery and Reinvestment Act, the next phase has begun on decommissioning the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site (SRS). February 23, 2011 Major Recovery Act Project Completed at Hanford: Two New Super Cells Go Into Service To Accept Contaminated Soil and Debris Months Ahead of Schedule and Millions Under Budget RICHLAND, WASH. - Two new super cells are going into service to expand disposal capacity for contaminated soil and debris at the Environmental Restoration Disposal Facility (ERDF), at the Department of Energy's (DOE) Hanford Site in southeastern Washington State. February 9, 2011

228

Development and test of a space-reactor-core heat pipe  

SciTech Connect

A heat pipe designed to meet the heat transfer requirements of a 100-kW/sub e/ space nuclear power system has been developed and tested. General design requirements for the device included an operating temperature of 1500/sup 0/K with an evaporator radial flux density of 100 w/cm/sup 2/. The total heat-pipe length of 2 m comprised an evaporator length of 0.3 m, a 1.2-m adiabatic section, and a condenser length of 0.5 m. A four-artery design employing screen arteries and distribution wicks was used with lithium serving as the working fluid. Molybdenum alloys were used for the screen materials and tube shell. Hafnium and zirconium gettering materials were used in connection with a pre-purified distilled lithium charge to ensure internal chemical compatibility. After initial performance verification, the 14.1-mm i.d. heat pipe was operated at 15 kW throughput at 1500/sup 0/K for 100 hours. No performance degradation was observed during the test.

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

1983-01-01T23:59:59.000Z

229

Deviation of the Test Program and Procedures for the 710 Critical Experiment Reactor Related to Changes in the core Material Volume Fractions  

SciTech Connect

This document describes a deviation from the "Test Program and Procedures for the 710 Critical Experiment Reactor Loading and Rod Calibrations," TM-63-1-702, which was made in accordance with ITS Standard Practice J80-81 on March 13, 1963. The deviation did not involve a significant change in the safety of the operation.

Sims, F.L.

1963-03-13T23:59:59.000Z

230

Production Test IP-358-AC: Replacement of carbon dioxide with nitrogen as a constituent of the K reactor atmosphere  

SciTech Connect

Compensation for the positive long-term reactivity transient associated with Hanford reactor may be accomplished in two ways: The addition of a poisonous material (rods, splines, etc.) to the reactor, or cooling the moderator by changing the gas composition. The objective of this study is to investigate the reactivity and temperature effects and the associated operating problems if any, resulting from the use of nitrogen instead of carbon dioxide as a constituent of the reactor atmosphere.

Bailey, G.F.; Benoliel, R.W.

1960-10-03T23:59:59.000Z

231

Lifetime Test of a Partial Model of a High-Temperature Gas-Cooled Reactor Helium-Helium Heat Exchanger  

Science Conference Proceedings (OSTI)

H. Design Codes and Life Prediction / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

Masaki Kitagawa; Hiroshi Hattori; Akira Ohtomo; Tetsuo Teramae; Junichi Hamanaka; Hiroshi Ukikusa

232

Temporal behavior of neutral particle fluxes in TFTR (Tokamak Fusion Test Reactor) neutral beam injectors  

SciTech Connect

Data from an E {parallel} B charge exchange neutral analyzer (CENA), which views down the axis of a neutral beamline through an aperture in the target chamber calorimeter of the TFTR neutral beam test facility, exhibit two curious effects. First, there is a turn-on transient lasting tens of milliseconds having a magnitude up to three times that of the steady-state level. Second, there is a 720 Hz, up to 20% peak-to-peak fluctuation persisting the entire pulse duration. The turn-on transient occurs as the neutralizer/ion source system reaches a new pressure equilibrium following the effective ion source gas throughput reduction by particle removal as ion beam. Widths of the transient are a function of the gas throughput into the ion source, decreasing as the gas supply rate is reduced. Heating of the neutalizer gas by the beam is assumed responsible, with gas temperature increasing as gas supply rate is decreased. At low gas supply rates, the transient is primarliy due to dynamic changes in the neutralizer line density and/or beam species composition. Light emission from the drift duct corroborate the CENA data. At high gas supply rates, dynamic changes in component divergence and/or spatial profiles of the source plasma are necessary to explain the observations. The 720 Hz fluctuation is attributed to a 3% peak-to-peak ripple of 720 Hz on the arc power supply amplified by the quadratic relationship between beam divergence and beam current. Tight collimation by CENA apertures cause it to accept a very small part of the ion source's velocity space, producing a signal linearly proportional to beam divergence. Estimated fluctuations in the peak power density delivered to the plasma under these conditions are a modest 3--8% peak to peak. The efffects of both phenomena on the injected neutral beam can be ameliorated by careful operion of the ion sources. 21 refs., 11 figs., 2 tabs.

Kamperschroer, J.H.; Gammel, G.M.; Roquemore, A.L.; Grisham, L.R.; Kugel, H.W.; Medley, S.S.; O' Connor, T.E.; Stevenson, T.N.; von Halle, A.; Williams, M.D.

1989-09-01T23:59:59.000Z

233

2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2012-02-01T23:59:59.000Z

234

2012 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2011 through October 31, 2012. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance issues Discussion of the facility’s environmental impacts During the 2012 permit year, approximately 183 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2013-02-01T23:59:59.000Z

235

Measurements of the radial structure and poloidal spectra of toroidal Alfven eigenmodes in the Tokamak Fusion Test Reactor  

SciTech Connect

Toroidal Alfven eigenmodes (TAE) have been excited by tangential neutral beam injection in the Tokamak Fusion Test Reactor (TFTR) [[ital Proceedings] [ital of] [ital the] [ital Thirteenth] [ital International] [ital Conference] [ital on] [ital Plasma] [ital Physics] [ital and] [ital Controlled] [ital Nuclear] [ital Fusion] [ital Research], 1990, Washington, D.C. (International Atomic Energy Agency, Vienna, 1990), Vol. I, p. 9]. Beam emission spectroscopy (BES) has been used to study the radial structure and the poloidal power spectra of these modes. Radial profiles show a global, standing wave structure with a node near [ital r]/[ital a]=0.6 and a maximum displacement of about 5--10 mm. The cross-phase profiles and the power spectra both imply that the mode is composed of a mixture of components with various poloidal and toroidal mode numbers, as expected for the TAE. Measurements of the poloidal mode spectrum via BES show good agreement with theoretical simulations performed by a nonvariational, kinetic magnetohydrodynamics stability code (NOVA[minus]K [Cheng, Phys. Rep. [bold 211], 1 (1992)]). In particular, the dominant harmonics in the poloidal spectrum obey the expected relation [ital m]+1/2[approx][ital q]([ital r])[ital n].

Durst, R.D.; Fonck, R.J. (University of Wisconsin, Madison, Wisconsin 53706 (United States)); Wong, K.L.; Cheng, C.Z.; Fredrickson, E.D.; Paul, S.F. (Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States))

1992-11-01T23:59:59.000Z

236

Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029  

SciTech Connect

This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

2013-09-01T23:59:59.000Z

237

Mapping of a reactor coolant effluent ground disposal test using an infrared imaging system and ground water potential and temperature measurements  

SciTech Connect

The concept of reactor effluent disposal to ground in infiltration trenches was proposed by Nelson and Alkire in 1963. At that time the available data indicated that radionuclide infiltration rates were probably adequate for trench disposal and that decontamination factors of 10 to 100 should be obtainable. Field tests at 100-F Area 1965 and 100-D Area 1967 have indicated that the infiltration rates are adequate and DF`s of from 2.5 for {sup 51}Cr to 7276 for {sup 65}Zn were obtained during the 100-D test. The purpose of this report is to present the results and interpretations of data from studies conducted over a reactor coolant effluent disposal test site. Data presented in this report were collected over the 100-C Area test in which a significant percentage of the reactor coolant effluent was disposed to an existing trench for a five-month period. Results of infrared thermal surveys and ground water temperature and potential measurements collected during this test are presented.

Eliason, J.R.

1969-04-10T23:59:59.000Z

238

Materials testing and development of functionally graded composite fuel cladding and piping for the Lead-Bismuth cooled nuclear reactor  

E-Print Network (OSTI)

This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite ...

Fray, Elliott Shepard

2013-01-01T23:59:59.000Z

239

Out-of-Reactor Corrosion Tests of Fuel Cladding Materials: Corrosion as a Function of Hydrogen Overpressure  

Science Conference Proceedings (OSTI)

EPRI has sponsored laboratory experiments to investigate whether an increased dissolved hydrogen (DH) level in the reactor coolant of pressurized water reactors (PWR) would result in increased hydrogen pickup (HPU) by the fuel cladding and spacer weld structure materials. This report documents exposure of clean, modern zirconium-based alloys for up to 730 days at three DH levels as well as exposure of Zircaloy 4 (Zry-4) specimens with different types of nickel contacts for 100 days at three DH ...

2013-11-27T23:59:59.000Z

240

Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel  

Science Conference Proceedings (OSTI)

High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

Fauzia, A. F.; Waris, A.; Novitrian [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, INDONESIA Jl. Ganesa 10 Bandung 40132 (Indonesia)

2010-06-22T23:59:59.000Z

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241

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

242

High poloidal beta equilibria in the Tokamak Fusion Test Reactor limited by a natural inboard poloidal field null  

DOE Green Energy (OSTI)

Recent operation of the Tokamak Fusion Test Reactor (TFTR) (Plasma Phys. Controlled Nucl. Fusion Research {bold 1}, 51 (1986)) has produced plasma equilibria with values of {Lambda}{equivalent to}{beta}{sub {ital p} eq}+{ital l}{sub {ital i}}/2 as large as 7, {epsilon}{beta}{sub {ital p} dia}{equivalent to}2{mu}{sub 0}{epsilon}{l angle}{ital p}{sub {perpendicular}}{r angle}/{l angle}{l angle}{ital B}{sub {ital p}}{r angle}{r angle}{sup 2} as large as 1.6, and Troyon normalized diamagnetic beta (Plasma Phys. Controlled Fusion {bold 26}, 209 (1984); Phys. Lett. {bold 110A}, 29 (1985)), {beta}{sub {ital N}dia}{equivalent to}10{sup 8}{l angle}{beta}{sub {ital t}{perpendicular}}{r angle}{ital aB}{sub 0}/{ital I}{sub {ital p}} as large as 4.7. When {epsilon}{beta}{sub {ital p} dia}{approx gt}1.25, a separatrix entered the vacuum chamber, producing a naturally diverted discharge that was sustained for many energy confinement times, {tau}{sub {ital E}}. The largest values of {epsilon}{beta}{sub {ital p}} and plasma stored energy were obtained when the plasma current was ramped down prior to neutral beam injection. The measured peak ion and electron temperatures were as large as 24 and 8.5 keV, respectively. Plasma stored energy in excess of 2.5 MJ and {tau}{sub {ital E}} greater than 130 msec were obtained. Confinement times of greater than 3 times that expected from L-mode predictions have been achieved. The fusion power gain {ital Q}{sub DD} reached a value of 1.3{times}10{sup {minus}3} in a discharge with {ital I}{sub {ital p}}=1 MA and {epsilon}{beta}{sub {ital p} dia}=0.85. A large, sustained negative loop voltage during the steady-state portion of the discharge indicates that a substantial noninductive component of {ital I}{sub {ital p}} exists in these plasmas. Transport code analysis indicates that the bootstrap current constitutes up to 65% of {ital I}{sub {ital p}}.

Sabbagh, S.A.; Gross, R.A.; Mauel, M.E.; Navratil, G.A. (Department of Applied Physics, Columbia University, New York, New York 10027 (USA)); Bell, M.G.; Bell, R.; Bitter, M.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Chance, M.S.; Efthimion, P.C.; Fredrickson, E.D.; Hatcher, R.; Hawryluk, R.J.; Hirshman, S.P.; Janos, A.C.; Jardin, S.C.; Jassby, D.L.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Okabayashi, M.; Park, H.K.; Ramsey, A.T.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Wieland, R.M.; Zarnstorff, M.C. (Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (USA)); Kesner, J.; Marmar, E.S.; Terry, J.L. (MIT Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (USA))

1991-08-01T23:59:59.000Z

243

HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2  

SciTech Connect

This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-630 Catch Tank System closure plan due to ongoing laboratory operations in the areas requiring closure actions. The TRA-604 laboratory components include the TRA-604 laboratory warm wastewater drain piping, undersink drains, subheaders, and the east TRA-604 laboratory drain header. Potentially contaminated surfaces located beneath the TRA-604 laboratory warm wastewater drain piping and beneath the island sinks located in Laboratories 126 and 128 (located in TRA-661) are also addressed in this closure plan. The TRA-604 laboratory components will be closed in accordance with the interim status requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, Subparts G and J. This closure plan presents the closure performance standards and the methods for achieving those standards.

KIRK WINTERHOLLER

2008-02-25T23:59:59.000Z

244

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

245

Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)  

SciTech Connect

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

2012-07-01T23:59:59.000Z

246

The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors  

SciTech Connect

A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

1987-10-01T23:59:59.000Z

247

Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thicknesses of 1.27 or 1.52 mm. Fuel core dimensions varied according to the overall plate thicknesses with a minimum clad thickness of 0.20 mm. Tested fuels included UAl/sub x/, UAl/sub 2/, U/sub 3/O/sub 8/, U/sub 3/SiAl, U/sub 3/Si, U/sub 3/Si/sub 1.5/, U/sub 3/Si/sub 2/, U/sub 3/SiCu, USi, U/sub 6/Fe, and U/sub 6/Mn/sub 1.3/ materials. Although most miniplates were made with low-enriched uranium (19.9%), some with medium-enriched uranium (40 to 45%), a few with high-enriched uranium (93%), and a few with depleted uranium (0.2 to 0.4%) were tested for comparison. These fuel materials were irradiated to burnups ranging from /approximately/27 to 98 at. % /sup 235/U depletion. Operation of the experiment, measurement of miniplate thickness as the irradiation progressed, ultimate shipment of the irradiated miniplates to various hot cells, and preliminary results are reported here. 18 refs., 12 figs., 7 tabs.

Senn, R.L.

1989-04-01T23:59:59.000Z

248

Technique for preparation of prototypic fast test reactor mixed-oxide fuel pins for ceramography and metallography  

Science Conference Proceedings (OSTI)

Irradiated liquid metal fast breeder reactor prototypic mixed-oxide fuel and cladding have been sectioned and examined in remote hot cell facilities for ceramographic and metallographic examinations. Metallography, shielded electron microprobe examinations, electron microscopy from replicas and autoradiography are routinely employed to obtain postirradiation data for statistical evaluation. Selection of preparatory techniques are based on fuel type, burnup, customer requirements and fuel pin condition.

Chastain, S.A.

1981-01-01T23:59:59.000Z

249

THE HOMOGENEOUS SUSPENSION REACTOR PROJECT  

SciTech Connect

The considerations which led to the study of a homogeneous suspension reactor are reviewed briefly. The characteristics of the KEMA Suspension Test Reactor (KSTR) are then summarized. (J.S.R.)

Went, J.J.

1963-02-01T23:59:59.000Z

250

RESULTS OF TESTS TO DEMONSTRATE A SIX-INCH DIAMETER COATER FOR PRODUCTION OF TRISO-COATED PARTICLES FOR ADVANCED GAS REACTOR EXPERIMENTS  

DOE Green Energy (OSTI)

The Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program includes a series of irradiation experiments in Idaho National Laboratory's (INL's) Advanced Test Reactor. TRISOcoated particles for the first AGR experiment, AGR-1, were produced at Oak Ridge National Laboratory (ORNL) in a twoinch diameter coater. A requirement of the NGNP/AGR Program is to produce coated particles for later experiments in coaters more representative of industrial scale. Toward this end, tests have been performed by Babcock and Wilcox (B&W) in a six-inch diameter coater. These tests are expected to lead to successful fabrication of particles for the second AGR experiment, AGR-2. While a thorough study of how coating parameters affect particle properties was not the goal of these tests, the test data obtained provides insight into process parameter/coated particle property relationships. Most relationships for the six-inch diameter coater followed trends found with the ORNL two-inch coater, in spite of differences in coater design and bed hydrodynamics. For example the key coating parameters affecting pyrocarbon anisotropy were coater temperature, coating gas fraction, total gas flow rate and kernel charge size. Anisotropy of the outer pyrolytic carbon (OPyC) layer also strongly correlates with coater differential pressure. In an effort to reduce the total particle fabrication run time, silicon carbide (SiC) was deposited with methyltrichlorosilane (MTS) concentrations up to 3 mol %. Using only hydrogen as the fluidizing gas, the high concentration MTS tests resulted in particles with lower than desired SiC densities. However when hydrogen was partially replaced with argon, high SiC densities were achieved with the high MTS gas fraction.

Douglas W. Marshall

2008-09-01T23:59:59.000Z

251

RESULTS OF TESTS TO DEMONSTRATE A SIX-INCH-DIAMETER COATER FOR PRODUCTION OF TRISO-COATED PARTICLES FOR ADVANCED GAS REACTOR EXPERIMENTS  

DOE Green Energy (OSTI)

The Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program includes a series of irradiation experiments in Idaho National Laboratory’s (INL’s) Advanced Test Reactor. TRISOcoated particles for the first AGR experiment, AGR-1, were produced at Oak Ridge National Laboratory (ORNL) in a two inch diameter coater. A requirement of the NGNP/AGR Program is to produce coated particles for later experiments in coaters more representative of industrial scale. Toward this end, tests have been performed by Babcock and Wilcox (B&W) in a six-inch diameter coater. These tests are expected to lead to successful fabrication of particles for the second AGR experiment, AGR-2. While a thorough study of how coating parameters affect particle properties was not the goal of these tests, the test data obtained provides insight into process parameter/coated particle property relationships. Most relationships for the six-inch diameter coater followed trends found with the ORNL two-inch coater, in spite of differences in coater design and bed hydrodynamics. For example the key coating parameters affecting pyrocarbon anisotropy were coater temperature, coating gas fraction, total gas flow rate and kernel charge size. Anisotropy of the outer pyrolytic carbon (OPyC) layer also strongly correlates with coater differential pressure. In an effort to reduce the total particle fabrication run time, silicon carbide (SiC) was deposited with methyltrichlorosilane (MTS) concentrations up to 3 mol %. Using only hydrogen as the fluidizing gas, the high concentration MTS tests resulted in particles with lower than desired SiC densities. However when hydrogen was partially replaced with argon, high SiC densities were achieved with the high MTS gas fraction.

Charles M Barnes

2008-09-01T23:59:59.000Z

252

Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

Freels, James D [ORNL; Jain, Prashant K [ORNL; Hobbs, Randy W [ORNL

2012-01-01T23:59:59.000Z

253

Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project  

SciTech Connect

The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

2000-04-01T23:59:59.000Z

254

Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project  

SciTech Connect

The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

Del Cul, G.D.

2000-06-07T23:59:59.000Z

255

Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors  

Science Conference Proceedings (OSTI)

A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000şC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

Dawn Scates

2010-10-01T23:59:59.000Z

256

Resumption of Transient Testing | Department of Energy  

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

that will generate additional quantities of clean, reliable, economical electricity with nuclear power reactors. The Transient Reactor Test Facility (TREAT) reactor at Idaho...

257

ACDOS1: A COMPUTER CODE TO CALCULATE DOSE RATES FROM NEUTRON ACTIVATION OF NEUTRAL BEAMLINES AND OTHER FUSION REACTOR COMPONENTS  

E-Print Network (OSTI)

BEAMLINES AND OTHER FUSION--REACTOR COMPONENTS Gregory S.Beamlines and Other Fusion-Reactor Compon­ ents By Gregoryin the Tokamak Fusion Test Reactor Test Cell", Nucl.

Keney, G.S.

2010-01-01T23:59:59.000Z

258

Nuclear Reactors  

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

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

259

Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments  

DOE Green Energy (OSTI)

In fission and fusion reactor environments stainless steels generate significant amounts of helium and hydrogen by transmutation. The primary sources of helium are boron and nickel, interacting with both fast and especially thermal neutrons. Hydrogen arises primarily from fast neutron reactions, but is also introduced into steels at often much higher levels by other environmental processes. Although essentially all of the helium is retained in the steel, it is commonly assumed that most of the hydrogen is not retained. It now appears that under some circumstances, significant levels of hydrogen can be retained, especially when helium-nucleated cavities become a significant part of the microstructure. A variety of stainless steel specimens have been examined from various test reactors, PWRs and BWRs. These specimens were exposed to a wide range of neutron spectra with different thermal/fast neutron ratios. Pure nickel and pure iron have also been examined. It is shown that all major features of the retention of helium and hydrogen can be explained in terms of the composition, thermal/fast neutron ratio and the presence or absence of helium-nucleated cavities. In some cases, the hydrogen retention is very large and can exceed that generated by transmutation, with the additional hydrogen arising from either environmental sources and/or previously unidentified radioisotope sources that may come into operation at high neutron exposures.

Garner, Francis A.; Oliver, Brian M.; Greenwood, Lawrence R.; Edwards, Danny J.; Bruemmer, Stephen M.; Grossbeck, Martin L.

2002-03-31T23:59:59.000Z

260

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop  

SciTech Connect

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

McCulloch, R.W.; MacPherson, R.E.

1983-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report  

SciTech Connect

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

Hauptman, H.M.; Petro, J.N.; Jacobi, O. [and others

1995-04-01T23:59:59.000Z

262

Ensuring the Performance of Nuclear Reactor Pressure Vessels for ...  

Science Conference Proceedings (OSTI)

The Light Water Reactor Sustainability Program is a collaborative program ... and in situ Mechanical Test Methods in the US Fusion Reactor Materials Program.

263

H. R. 1001: A Bill to authorize appropriations for the Reduced Enrichment Research and Test Reactors Program of the Department of Energy. Introduced in the House of Representatives, One Hundred Third Congress, First Session, February 18, 1993  

SciTech Connect

This Act may be cited as the [open quotes]Bomb-Grade Uranium Export Substitution Act of 1993[close quotes]. The purpose of this Bill is to authorize appropriations for the Reduced Enrichment Research and Test Reactors Program of the Department of Energy. This document presents Congressional findings and a statement of authorization of appropriations.

1993-01-01T23:59:59.000Z

264

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

265

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Daniels, F.

1959-10-27T23:59:59.000Z

266

CONVECTION REACTOR  

DOE Patents (OSTI)

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Hammond, R.P.; King, L.D.P.

1960-03-22T23:59:59.000Z

267

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, FEBRUARY 1961  

SciTech Connect

Design, development, and testing efforts were continued on BORAX-V, EBR- I, EBR-II, EBWR, JUGGERNAUT ZPRIII, ZPR-VI, and ZPR-W. An evaluation program is outlined for Pebble Bed Reactor designs. Fast and thermal reactor safety studies were conducted. Experimental and theoretical studies in applied nuclear and reactor physics are dsscribed. Developments made in reactor components, fuels, and materials are discussed. Heat engineering studies were conducted on steam separation, and velocity and void distributions in two-phase systems. Fluidization and fluoride volatility separation, and chemical-metallurgical separation processes were studied. Advanced reactor concepts that were discusssd includsed. Basic Radiation Effects Beactor, Biogeonuclear Reactor, Fast Reactor Test Facility, compact high-power density fast reactors, AHFR hydraulic test loop, Packed Bed Reactor, and direct conversion. (For preceding period see ANL- 6328.) (B.O.G.)

1961-03-15T23:59:59.000Z

268

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01T23:59:59.000Z

269

EA-0813; Environmental Assessment and (FONSI) The Tokamak Fusion Test Reactor Decontamination and Decommissioning Project and The Tokamak Physics Experiment at the PPPL  

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

13; Environmental Assessment and (FONSI) The Tokamak Fusion 13; Environmental Assessment and (FONSI) The Tokamak Fusion Test Reactor Decontamination and Decommissioning Project and The Tokamak Physics Experiment at the PPPL Table of Contents EXECUTIVE SUMMARY ACRONYMS Glossary of Radiological Terms SCIENTIFIC NOTATION 1.0 PURPOSE AND NEED FOR THE PROPOSED ACTIONS 1.1 TFTR D&D Project 1.2 TPX Project 1.3 Scope of Document 1.4 Local Community Relations Program 1.5 References 2.0 DESCRIPTION OF THE PROPOSED ACTIONS AND ALTERNATIVES 2.1 TFTR D&D Project 2.2 TPX Project 2.3 Environmental Monitoring 2.4 References 3.0 DESCRIPTION OF THE AFFECTED ENVIRONMENT 3.1 PPPL Proposed Site 3.2 ORR Alternative Site 3.3 References 4.0 ENVIRONMENTAL CONSEQUENCES OF THE PROPOSED ACTIONS AND ALTERNATIVES 4.1 TFTR D&D Project 4.1.1 Impacts of Normal D&D Operations

270

2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond  

Science Conference Proceedings (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Discussion of the facility’s environmental impacts During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

mike lewis

2011-02-01T23:59:59.000Z

271

A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94  

SciTech Connect

Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA).

B. R. Orr (USGS)

1999-11-01T23:59:59.000Z

272

TEST  

Science Conference Proceedings (OSTI)

This is an abstract. TEST Lorem ipsum dolor sit amet, consectetur adipiscing elit. Cras lacinia dui et est venenatis lacinia. Vestibulum lacus dolor, adipiscing id mattis sit amet, ultricies sed purus. Nulla consectetur aliquet feugiat. Maecenas ips

273

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

274

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Fraas, A.P.; Mills, C.B.

1961-11-21T23:59:59.000Z

275

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

276

Tokamak Fusion Test Reactor experiment  

SciTech Connect

A brief review of the TFTR is given in terms of the physical size of the experiment in relation to existing and future tokamaks. Some break-even criteria are mentioned. (MOW)

Furth, H.P.

1976-01-01T23:59:59.000Z

277

Fusion Test Facilities John Sheffield  

E-Print Network (OSTI)

flexing tests - Testing nuclear fuel assemblies to meltdown--PHEBUS reactor #12;#12;Released on February REACTOR--CADARACHE · Purpose: studies of hypothetical accidents in pressurized water reactors · Type: pool.78% · The reactor was transformed into a miniature PWR (scale 1/5000) for the program Phébus PFF, a study

278

DESTRUCTIVE EXAMINATION OF 3-CYCLE LWR (LIGHT WATER REACTOR) FUEL RODS FROM TURKEY POINT UNIT 3 FOR THE CLIMAX - SPENT FUEL TEST  

DOE Green Energy (OSTI)

The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reator fuel rods with similar burnups (28 GWd/MTU) and operating histories.

ATKIN SD

1981-06-01T23:59:59.000Z

279

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

280

Nuclear Reactor Severe Accident Experiments  

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

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

Note: This page contains sample records for the topic "test reactor hwctr" 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

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11T23:59:59.000Z

282

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

283

Reactor Materials  

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

The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...

284

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

285

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

286

REACTOR SHIELD  

DOE Patents (OSTI)

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17T23:59:59.000Z

287

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

288

REACTOR GROUT THERMAL PROPERTIES  

DOE Green Energy (OSTI)

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

289

Coupled thermohydromechanical analysis of a heater test in unsaturated clay and fractured rock at Kamaishi Mine  

E-Print Network (OSTI)

Instrumentation. Power reactor and nuclear fiel developmentexperiment area. Power reactor and nuclear fbel developmentHydraulic tests. Power reactor and nuclear fiel development

Rutqvist, J.

2011-01-01T23:59:59.000Z

290

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

291

Brookhaven Medical Research Reactor  

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

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

292

NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01T23:59:59.000Z

293

Materials Reliability Program: Testing and Evaluation of Reactor Pressure Vessel Steel Plate Heat JRQ to Assess Through-Wall Attenua tion of Radiation Embrittlement (MRP-243)  

Science Conference Proceedings (OSTI)

The change in neutron energy spectrum through the wall of a reactor pressure vessel (RPV) requires the use of an exposure parameter or metric for assessing radiation embrittlement. This report looks at experimental fracture toughness and Charpy V-notch (CVN) data generated in a special International Atomic Energy Agency (IAEA) experiment designed to simulate an RPV wall of 190 mm thickness. These experimental data are compared with the current exposure metric of displacements per atom (dpa) coupled with ...

2008-12-23T23:59:59.000Z

294

Materials Reliability Program: Testing and Evaluation of Two Reactor Pressure Vessel Steels Irradiated to Assess Through-Wall Attenu ation of Radiation Embrittlement (MRP-203)  

Science Conference Proceedings (OSTI)

The change in neutron energy spectrum through the wall of a reactor pressure vessel (RPV) requires the use of an exposure parameter or metric for assessing radiation embrittlement. This report looks at experimental fracture toughness and Charpy V-notch data generated in a special International Atomic Energy Agency (IAEA) experiment designed to simulate an RPV wall of 180-mm thickness. These experimental data are compared with the current exposure metric of displacements per atom (dpa) coupled with an emb...

2006-10-04T23:59:59.000Z

295

POWER REACTOR  

DOE Patents (OSTI)

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Zinn, W.H.

1958-07-01T23:59:59.000Z

296

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

297

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

298

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

299

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Wigner, E.P.; Young, G.J.

1958-10-14T23:59:59.000Z

300

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

302

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

303

Coupled Reactor Kinetics and Heat Transfer Model for Heat Pipe Cooled Reactors  

SciTech Connect

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). The paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities.

WRIGHT,STEVEN A.; HOUTS,MICHAEL

2000-11-22T23:59:59.000Z

304

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

305

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

306

Neutronic reactor  

DOE Patents (OSTI)

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

307

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

308

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

309

NEUTRONIC REACTOR  

DOE Patents (OSTI)

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27T23:59:59.000Z

310

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

311

The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases  

SciTech Connect

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest status and plans are presented.

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01T23:59:59.000Z

312

Results of charpy V-notch impact testing of structural steel specimens irradiated at {approximately}30{degrees}C to 1 x 10{sup 16} neutrons/cm{sup 2} in a commercial reactor cavity  

SciTech Connect

A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at {approximately} 30{degrees}C ({approximately} 85{degrees}F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10{sup 16} neutrons/cm{sup 2} (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was {approximately} 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of {approximately} 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications.

Iskander, S.K.; Stoller, R.E.

1997-04-01T23:59:59.000Z

313

ACDOS1: A COMPUTER CODE TO CALCULATE DOSE RATES FROM NEUTRON ACTIVATION OF NEUTRAL BEAMLINES AND OTHER FUSION REACTOR COMPONENTS  

E-Print Network (OSTI)

the Tokamak Fusion Test Reactor Test Cell", Nucl. Technol.BEAMLINES AND OTHER FUSION--REACTOR COMPONENTS Gregory S.Beamlines and Other Fusion-Reactor Compon­ ents By Gregory

Keney, G.S.

2010-01-01T23:59:59.000Z

314

REACTOR UNLOADING  

DOE Patents (OSTI)

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Leverett, M.C.

1958-02-18T23:59:59.000Z

315

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

316

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

317

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

318

REACTOR CONTROL  

DOE Patents (OSTI)

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Ruano, W.J.

1957-12-10T23:59:59.000Z

319

Research and Medical Isotope Reactor Supply | Y-12 National Security...  

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

Research and Medical ... Research and Medical Isotope Reactor Supply Our goal is to fuel research and test reactors with low-enriched uranium. Y-12 tops the short list of the...

320

Program for alloy development for irradiation performance in fusion reactors  

SciTech Connect

The use of fission reactors as irradiation test facilities for structural materials for a fusion environment is discussed. A comparison is made of displacement damage and helium production in fast fission and fusion reactors for stainless steel. (MOW)

Stiegler, J.O.; Reuther, T.C.

1977-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

Historical Photographs: Nevada Test Site  

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

Nevada Test Site Small Image 1. A nuclear reactor sitting on a test cell pad prior to preliminary tests at the Nevada Test Site (circa 1968). This Phoebus 2 design was part of...

322

Light Water Reactors Technology Development - Nuclear Reactors  

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

Light Water Reactors Light Water 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 Argonne's Nuclear Science and Technology Legacy

323

Experimental Breeder Reactor I Preservation Plan  

SciTech Connect

Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

Julie Braun

2006-10-01T23:59:59.000Z

324

Water simulation of sodium reactors  

Science Conference Proceedings (OSTI)

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system.

Grewal, S.S.; Gluekler, E.L.

1981-06-28T23:59:59.000Z

325

Hexagonal-geometry fast-reactor nodal modeling  

Science Conference Proceedings (OSTI)

Several nodal methods have been developed for simulating power distributions in thermal reactors, and have been tested for applicability to fast reactor problems. The testing was performed in rectangular geometry. Since fast reactor configurations typically use hexagonal assemblies, the most promising of the techniques tested was extended to hexagonal geometry and applied to a range of test cases. The approach selected was the generalized coarse mesh procedure (GCMDT) using different interpolation parameters in different zones.

Caro, E.; Becker, M.

1982-01-01T23:59:59.000Z

326

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

327

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

328

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

329

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements  

SciTech Connect

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17T23:59:59.000Z

330

Sodium Reaction Experimental Test Facility (SRETF) - Nuclear...  

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

Form Modeling Departments Engineering Analysis Nuclear Systems Analysis Research & Test Reactor Nonproliferation and National Security Detection & Diagnostic Systems...

331

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

332

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network (OSTI)

For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR;Equipment Layout #12;Modular Pebble Bed Reactor Thermal Power 250 MW Core Height 10.0 m Core Diameter 3.5 m · License by Test · Expert I&C System - Hands free operation #12;MIT MPBR Specifications Thermal Power 250

333

Nuclear Reactor Accidents  

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

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

334

Principles of Reactor Physics  

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

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

335

Experience in the Testing and Application of the EPRI DfD Process: Decontamination for Decommissioning of Reactor Coolant Systems and Plant Components  

Science Conference Proceedings (OSTI)

This report provides a vendor's perspective on the application of the EPRI DfD (Decontamination for Decommissioning) process. It includes a summary of in-house testing and initial field applications of the process, focusing on lessons learned and the application methodology.

1999-11-08T23:59:59.000Z

336

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

337

Relicensing of the MIT Research Reactor  

SciTech Connect

The Massachusetts Institute of Technology (MIT) Research Reactor (MITR) is owned and operated by MIT, a nonprofit university. The current reactor, MITR-II, is a 5-MW, light water-cooled and heavy water-moderated reactor that uses materials test reactor-type fuel. Documents supporting application to the U.S. Nuclear Regulatory Commission (NRC) for relicensing of MITR were submitted in July 1999. A power upgrade from 5 to 6 MW was also requested. The relicensed reactor (MITR-III) will be the third reactor operated by MIT. This paper describes MITR-I and MITR-II, and design options considered for MITR-III. Selected problems addressed during the relicensing studies are also described, namely core tank aging evaluation, neutronic analysis, thermal-hydraulic analysis, and step reactivity insertion analysis.

Lin-Wen Hu; John A. Bernard; Susan Tucker

2000-06-04T23:59:59.000Z

338

The LEU conversion status of U.S. Research Reactors.  

SciTech Connect

This paper summarizes the conversion status of US research and test reactors and estimates uranium densities needed to convert reactors with power levels 21 MW from HEU ({ge} 20% U-235) to LEU (<20% U-235) fuels. Detailed conversion studies for each reactor need to be completed in order to establish the feasibility of using LEU fuels.

Matos, J. E.

1997-11-14T23:59:59.000Z

339

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect

Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.

Schmidt, T.R. [Sandia National Labs., Albuquerque, NM (United States); Reuscher, J.A. [Texas A& M Univ., College Station, TX (United States)

1994-10-01T23:59:59.000Z

340

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

Engineering activities at the MIT research reactor in support of power reactor technology  

SciTech Connect

The Massachusetts Institute of Technology (MIT) research reactor (MITR-II) is a 5-MW(thermal) light-water-cooled and-moderated reactor (LWR) with in-core neutron and gamma dose rates that closely approximate those in current LWRs. Compact in-pile loops that simulate pressurized water reactor (PWR) and boiling water reactor (BWR) thermal hydraulics and coolant chemistry have been designed for installation in the MITR-II. A PWR loop has been completed and is currently operating in the reactor. A BWR loop is under construction, and an in-pile facility for irradiation-assisted stress corrosion crack (IASCC) testing is being designed. Another major area of research and on-line testing is the closed-loop, nonlinear, digital control of various reactor parameters, including the power level, temperature, and net energy production.

Harling, O.K.; Bernard, J.A.; Driscoll, M.J.; Kohse, G.E.; Ballinger, R.G.

1989-01-01T23:59:59.000Z

342

LOFT facility and test program  

SciTech Connect

The Loss-of-Fluid Test (LOFT) test facility, program objectives, and the experiments planned are described. The LOFT facility is related to the smaller Semiscale facility and the larger commercial pressurized water reactors. The fact that LOFT is a computer model assessment tool rather than a demonstration test is emphasized. Various types of reactor safety experiments planned through 1983 are presented.

McPherson, G.D.

1979-11-01T23:59:59.000Z

343

Control rod assembly for liquid metal fast breeder reactors  

SciTech Connect

This standard establishes the requirements for fabrication, testing, and inspection of control rod assemblies for use in liquid metal fast breeder reactors.

1978-09-08T23:59:59.000Z

344

Reactor safety method  

DOE Patents (OSTI)

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

345

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

346

INEEL/EXT-01-01623 MODULAR PEBBLE-BED REACTOR PROJECT  

E-Print Network (OSTI)

in the early 1990s. Fuel compacts were irradiated at the High Flux Isotope Reactor (HFIR) and the Advanced Test

347

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

348

Page not found | Department of Energy  

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

11 - 6620 of 28,905 results. 11 - 6620 of 28,905 results. Download CX-000504: Categorical Exclusion Determination Deactivation of 235-F Complex CX(s) Applied: B1.28 Date: 07/28/2009 Location(s): Aiken, South Carolina Office(s): Environmental Management, Savannah River Operations Office http://energy.gov/nepa/downloads/cx-000504-categorical-exclusion-determination Download CX-000509: Categorical Exclusion Determination Heavy Water Component Test Reactor (HWCTR) Well Installation CX(s) Applied: B3.1 Date: 08/05/2009 Location(s): Aiken, South Carolina Office(s): Environmental Management, Savannah River Operations Office http://energy.gov/nepa/downloads/cx-000509-categorical-exclusion-determination Download CX-000508: Categorical Exclusion Determination Electrical Isolation of Building 710-B

349

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

350

Transport Reactor Facility  

SciTech Connect

The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.

Berry, D.A.; Shoemaker, S.A.

1996-12-31T23:59:59.000Z

351

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, SEPTEMBER 1961  

SciTech Connect

BS>Data from examination of blade-type control rods which were used in BORAX are discussed. Operation and maintenance of EBWR is outlined. In work on Borax V, modifications for easier installation of reactor and components is outlined followed by discussion of superheat fuel element development, and fabrication of various reactor components. Borax reactor design is also reported along with information on development and testing. In research on sodiumcooled reactors, activities are summarized in the LPR III and LPR IV programs along with developmental work on EBR I and II. Studies on reactor safety are reported and activities in a program of nuclear technology and general support are outlined. (J.R.D.)

1961-10-15T23:59:59.000Z

352

THE EXPERIMENTAL BERYLLIUM OXIDE REACTOR. MARITIME GAS-COOLED REACTOR PROGRAM  

SciTech Connect

LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)

Moore, W.C.

1961-07-01T23:59:59.000Z

353

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...

354

Attrition reactor system  

DOE Patents (OSTI)

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

1993-01-01T23:59:59.000Z

355

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

356

Reactor Sharing Program  

Science Conference Proceedings (OSTI)

Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

Vernetson, W.G.

1993-01-01T23:59:59.000Z

357

Attrition reactor system  

DOE Patents (OSTI)

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Scott, C.D.; Davison, B.H.

1993-09-28T23:59:59.000Z

358

NEUTRONIC REACTOR POWER PLANT  

DOE Patents (OSTI)

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Metcalf, H.E.

1962-12-25T23:59:59.000Z

359

High solids fermentation reactor  

DOE Patents (OSTI)

A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

1993-01-01T23:59:59.000Z

360

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

FAST NEUTRON REACTOR  

DOE Patents (OSTI)

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

Soodak, H.; Wigner, E.P.

1961-07-25T23:59:59.000Z

362

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01T23:59:59.000Z

363

Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide  

Science Conference Proceedings (OSTI)

This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

2012-09-01T23:59:59.000Z

364

Prospects for Tokamak Fusion Reactors  

SciTech Connect

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

Sheffield, J.; Galambos, J.

1995-04-01T23:59:59.000Z

365

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

366

Nuclear reactor overflow line  

DOE Patents (OSTI)

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

367

Reactor vessel support system  

DOE Patents (OSTI)

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

368

Spinning fluids reactor  

SciTech Connect

A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

Miller, Jan D; Hupka, Jan; Aranowski, Robert

2012-11-20T23:59:59.000Z

369

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Cao, Jun

2011-01-01T23:59:59.000Z

370

Reactor water cleanup system  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

371

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Jun Cao

2011-01-12T23:59:59.000Z

372

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

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

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

373

Savannah River Site Removes Dome, Opening Reactor for Recovery Act  

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

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Savannah River Site this month after they safely removed its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the workers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning More Documents & Publications Recovery Act Changes Hanford Skyline with Explosive Demolitions Recovery Act Workers Add Time Capsule Before Sealing Reactor for Hundreds

374

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

375

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

376

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

377

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)  

SciTech Connect

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

GERBER MS

2009-04-28T23:59:59.000Z

378

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

379

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 reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

Spinrad, B.I.

1960-01-12T23:59:59.000Z

380

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Diebold, J.P.; Scahill, J.W.

1995-05-09T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" 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

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

382

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

383

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Isaksson, J.

1996-03-19T23:59:59.000Z

384

Tokamak reactor first wall  

DOE Patents (OSTI)

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

385

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

King, L.D.P.

1959-09-01T23:59:59.000Z

386

Fusion-reactor blanket and coolant material compatibility  

Science Conference Proceedings (OSTI)

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials.

Jeppson, D.W.; Keough, R.F.

1981-01-01T23:59:59.000Z

387

REACTOR DEVELOPMENT QUARTERLY PROGRESS REPORT  

DOE Green Energy (OSTI)

The design and construction program of the Bolling Experimental Reactor is reviewed. A number of preliminary experinnents were performed with Borax-II at pressures between 75 and 300 psi. The most corrosion-resistant U-Zr--Nb alloy developed so far is produced by heating in a vacuum to the gamma phase, quenching, and aging for 2 hr at 400 deg C. Special attention is given to the removal of H/sub 2/ from the material. Unclad and unirradiated samples of U--Nb and U--Nb--Zr alloys were corrosion tested in H/sub 2/O. Corrosion rates were also measured under irradiation conditions in CP-5. Elongation measurements of irradiated wrought and cast U--Zr material suggested no way for treating the wrought fuel so that stability comparable to the cast material could be obtained. Natural circulation boiling density tests at 600 psia were made in order to determine the effects of channel cross section and subcooling on the steam void fraction. Results of autoclave and dynamic corrosion studies of 2-S Al in H/sub 2/O are reported. These results include the testing of Ni-clad samples. A large number of criticality calculations were performed for the EBR-Il and the PBR. The solubility of ThO/sub 2/ pellets containing various concentrations of U/sub 3/ O/sub 8/ was tested in water at 316 deg C for periods of 672 to 744 hr. None of the samples disintegrated, although at least one sample developed cracks. Solutions of reactor kinetic equations were attempted for the purpose of studying transients in reactors with lifetimes of 7 x 10/sup -9/ 10/sup -7/, and 6 x 10/ sup -5/ sec. Ignition experiments were performed on Th, Cu, Al, Fe, Mg, Zr, and fluorothene when contacted with fluorides. Except for Zr and fluorothene, the materials did not ignite. (C.H.)

None

1955-01-31T23:59:59.000Z

388

Advances in Mechanical Testing - Programmaster.org  

Science Conference Proceedings (OSTI)

Feb 15, 2010 ... Testing of reactor irradiated materials for nuclear applications (fission .... Laboratory; 2Y-12 National Security Complex; 3University of Idaho

389

Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator...  

National Nuclear Security Administration (NNSA)

Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...

390

Production reactor characteristics  

SciTech Connect

Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.

Thiessen, C.W.; Hootman, H.E.

1990-01-01T23:59:59.000Z

391

Advanced converter reactors  

SciTech Connect

Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.

Kasten, P.R.

1979-01-01T23:59:59.000Z

392

NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Treshow, M.

1959-02-10T23:59:59.000Z

393

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Treshow, M.

1961-05-01T23:59:59.000Z

394

The Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

395

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

396

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

397

Radiation effects in materials for fusion reactors  

DOE Green Energy (OSTI)

The 14-MeV neutrons produced in a fusion reactor result in different irradiation damage than the equivalent fluence in a fast breeded reactor, not only because of the higher defect generation rate, but because of the production of significant concentrations of helium and hydrogen. Although no fusion test reactor exists, the effects of combined displacement damage plus helium can be studied in mixed-spectrum fission reactors for alloys containing nickel (e.g., austenitic stainless steels). The presence of helium appears to modify vacancy and interstitial recombination such that microstructural development in alloys differs between the fusion and fission reactor environments. Since mechanical properties of alloys are related to the microstructure, the simultaneous production of helium and displacement damage impacts upon key design properties such as tensile, fatigue, creep, an crack growth. Through an understanding of the basic phenomena occurring during irradiation and the relationships between microstructure and properties, alloys can be tailored to minimize radiation-induced swelling and improve mechanical properties in fusion reactor service.

Scott, J.L.; Grossbeck, M.L.; Maziasz, P.J.

1981-01-01T23:59:59.000Z

398

Test Automation Test Automation  

E-Print Network (OSTI)

Test Automation Test Automation Mohammad Mousavi Eindhoven University of Technology, The Netherlands Software Testing 2013 Mousavi: Test Automation #12;Test Automation Outline Test Automation Mousavi: Test Automation #12;Test Automation Why? Challenges of Manual Testing Test-case design: Choosing inputs

Mousavi, Mohammad

399

Global estimation of potential unreported plutonium in thermal research reactors  

SciTech Connect

As of November, 1993, 303 research reactors (research, test, training, prototype, and electricity producing) were operational worldwide; 155 of these were in non-nuclear weapon states. Of these 155 research reactors, 80 are thermal reactors that have a power rating of 1 MW(th) or greater and could be utilized to produce plutonium. A previously published study on the unreported plutonium production of six research reactors indicates that a minimum reactor power of 40 MW (th) is required to make a significant quantity (SQ), 8 kg, of fissile plutonium per year by unreported irradiations. As part of the Global Nuclear Material Control Model effort, we determined an upper bound on the maximum possible quantity of plutonium that could be produced by the 80 thermal research reactors in the non-nuclear weapon states (NNWS). We estimate that in one year a maximum of roughly one quarter of a metric ton (250 kg) of plutonium could be produced in these 80 NNWS thermal research reactors based on their reported power output. We have calculated the quantity of plutonium and the number of years that would be required to produce an SQ of plutonium in the 80 thermal research reactors and aggregated by NNWS. A safeguards approach for multiple thermal research reactors that can produce less than 1 SQ per year should be conducted in association with further developing a safeguards and design information reverification approach for states that have multiple research reactors.

Dreicer, J.S.; Rutherford, D.A.

1996-09-01T23:59:59.000Z

400

Nuclear reactor control column  

DOE Patents (OSTI)

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor hwctr" from the National Library of EnergyBeta (NLEBeta).
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401

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

402

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

403

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

Hopkins, R.J.; Land, J.T.; Misvel, M.C.

1994-06-07T23:59:59.000Z

404

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15T23:59:59.000Z

405

Spherical torus fusion reactor  

DOE Patents (OSTI)

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

406

CONTROL FOR NEUTRONIC REACTOR  

DOE Patents (OSTI)

S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

Lichtenberger, H.V.; Cameron, R.A.

1959-03-31T23:59:59.000Z

407

Fast Breeder Reactor studies  

Science Conference Proceedings (OSTI)

This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

1980-07-01T23:59:59.000Z

408

Microfluidic electrochemical reactors  

DOE Patents (OSTI)

A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

2011-03-22T23:59:59.000Z

409

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

410

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

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 in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

2009-03-01T23:59:59.000Z

411

Gas Test Loop Booster Fuel Hydraulic Testing  

SciTech Connect

The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

Gas Test Loop Hydraulic Testing Staff

2006-09-01T23:59:59.000Z

412

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants  

Science Conference Proceedings (OSTI)

Technical Paper / Second Seminar on Accelerated Testing of Materials in Spent Nuclear Fuel and High-Level Waste Storage Systems / Fission Reactors

Aung Tharn Daing; Myung-Hyun Kim

413

Gas-Cooled Fast Reactor Program. Annual progress report for period ending December 31, 1979  

SciTech Connect

Information on the GCFR reactor is presented concerning the Core Flow Test Loop; shielding and physics; pressure vessel and closure studies; and irradiation program.

Gat, U.; Kasten, P.R.

1980-11-01T23:59:59.000Z

414

Reactor Thermal-Hydraulics  

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

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

415

Reactor hot spot analysis  

SciTech Connect

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

416

THERMAL HYDRAULIC ANALYSIS OF A GAS TEST LOOP SYSTEM  

Science Conference Proceedings (OSTI)

This paper discusses thermal hydraulic calculations for a Gas Test Loop (GTL) system designed to provide a high intensity fast-flux irradiation environment for testing fuels and materials for advanced concept nuclear reactors. To assess the performance of candidate reactor fuels, these fuels must be irradiated under actual fast reactor flux conditions and operating environments, preferably in an existing irradiation facility [1]. Potential users of the GTL include the Generation IV Reactor Program, the Advanced Fuel Cycle Initiative and Space Nuclear Programs.

Donna Post Guillen; James E. Fisher

2005-11-01T23:59:59.000Z

417

Molten metal reactors  

SciTech Connect

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

418

Compact power reactor  

DOE Patents (OSTI)

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Wetch, Joseph R. (Woodland Hills, CA); Dieckamp, Herman M. (Canoga Park, CA); Wilson, Lewis A. (Canoga Park, CA)

1978-01-01T23:59:59.000Z

419

Solar Thermal Reactor Materials Characterization  

DOE Green Energy (OSTI)

Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor</