Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

DOE PAGES

Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; ...

2015-10-26

This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t 1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Theoretical Estimate of Maximum Possible Nuclear Explosion

DOE R&D Accomplishments Database

Bethe, H. A.

1950-01-31

The maximum nuclear accident which could occur in a Na-cooled, Be moderated, Pu and power producing reactor is estimated theoretically. (T.R.H.) 2O82 Results of nuclear calculations for a variety of compositions of fast, heterogeneous, sodium-cooled, U-235-fueled, plutonium- and power-producing reactors are reported. Core compositions typical of plate-, pin-, or wire-type fuel elements and with uranium as metal, alloy, and oxide were considered. These compositions included atom ratios in the following range: U-23B to U-235 from 2 to 8; sodium to U-235 from 1.5 to 12; iron to U-235 from 5 to 18; and vanadium to U-235 from 11 to 33. Calculations were performed to determine the effect of lead and iron reflectors between the core and blanket. Both natural and depleted uranium were evaluated as the blanket fertile material. Reactors were compared on a basis of conversion ratio, specific power, and the product of both. The calculated results are in general agreement with the experimental results from fast reactor assemblies. An analysis of the effect of new cross-section values as they became available is included. (auth)

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

NASA Astrophysics Data System (ADS)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

76 FR 16458 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-23

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR); Notice of Meeting The ACRS Subcommittee on U.S. EPR... of the Safety Evaluation Report (SER) with open items associated with U.S. EPR Design Control...

Monitoring bird population trends in North America

USGS Publications Warehouse

Robbins, C.S.; Bystrak, D.; Geissler, P.H.; Purroy, F.J.

1983-01-01

Se ofrece un nuevo metodo para computar las oscilaciones demograficas de las aves a lo largo de los anos. Con los datos suministrados por el proyecto 'Aves nidificantes en Norteamerica' , se indican en la Tabla 1 las pautas de cambio numerico de una serie seleccionada de aves holarticas.

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2010 CFR

2010-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2011 CFR

2011-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

76 FR 4739 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-26

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR); Notice of Meeting The ACRS Subcommittee on U.S. EPR... 6 and Chapter 15 of the U.S. EPR Safety Evaluation Report (SER) with Open Items. The Subcommittee...

Lengua nacional: Fichas de trabajo 1. Metodo de lectura y escritura (National Language: Workbook 1. A Method of Reading and Writing).

ERIC Educational Resources Information Center

Peniche Leger, Maria Elena, Ed.

This consumable, graded workbook can be used for exercises, tests, and individualized learning. Each level contains 30 units divided into four groups of exercises: reading analysis, grammar, composition, and spelling. Ten general tests are also included. For the accompanying reader, see FL 004 047. (Author/SK)

Ecological and toxicological aspects of the partial meltdown of the Chernobyl nuclear power plant reactor

USGS Publications Warehouse

Eisler, Ronald; Hoffman, David J.; Rattner, Barnett A.; Burton, G. Allen; Cairns, John

1995-01-01

the partial meltdown of the 1000-MW reactor at Chernobyl, Ukraine, on April 26, 1986, released large amounts of radiocesium and other radionuclides into the environment, causing widespread radioactive contamination of Europe and the former Soviet Union.1-7 At least 3,000,000 trillion becquerels (TBq) were released from the fuel during the accident (Table 24.1), dwarfing, by orders of magnitude, radiation released from other highly publicized reactor accidents at Windscale (U.K.) and three-Mile Island (U.S.)3,8 The Chernobyl accident happened while a test was being conducted during a normal scheduled shutdown and is attributed mainly to human error.3

77 FR 36581 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-19

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor (US-APWR) will hold a meeting on July 9-10, 2012, Room T-2B3, 11545...

Use and Analysis of Finite Element Methods for Problems of Solid Mechanics and Fracture

DTIC Science & Technology

1993-01-19

improve global accuracy and convergence rates, specific activities such as mesh refinement may be undertaken local to the crack tips. A criticism ...Sciences, Centro Internacional de Metodos Numnericos en Ingenieria, Barcelona, 1992 11 1 |l!avacek, J Rosenberg, A E Beagles and J R Whiteman

76 FR 77021 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-09

... searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2), NRC-2008-0231...://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: (301) 415-1035...

76 FR 71608 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-18

... searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2), NRC-2008-0231...://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: (301) 415-1035...

76 FR 75566 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-02

....regulations.gov by searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2... available at http://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone...

76 FR 72725 - Notice of Availability of Combined License Applications

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-25

... searching on Docket ID NRC-2008-0170 (William States Lee III Nuclear Station Units 1 and 2), NRC-2008-0231...://www.nrc.gov/reactors/new-reactors/col.html . FOR FURTHER INFORMATION CONTACT: Donald Habib, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: (301) 415-1035...

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

Health physics aspects of advanced reactor licensing reviews

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

Integral cross section measurement of the U 235 ( n , n ' ) U 235 m reaction in a pulsed reactor

DOE PAGES

Bélier, G.; Bond, E. M.; Vieira, D. J.; ...

2015-04-08

The integral measurement of the neutron inelastic cross section leading to the 26-minute half-life 235mU isomer in a fission-like neutron spectrum is presented. The experiment has been performed at a pulsed reactor, where the internal conversion decay of the isomer was measured using a dedicated electron detector after activation. The sample preparation, efficiency measurement, irradiation, radiochemistry purification, and isomer decay measurement will be presented. We determined the integral cross section for the ²³⁵U(n,n') 235mU reaction to be 1.00±0.13b. This result supports an evaluation performed with TALYS-1.4 code with respect to the isomer excitation as well as the total neutron inelasticmore » scattering cross section.« less

Detecting Dark Photons with Reactor Neutrino Experiments

NASA Astrophysics Data System (ADS)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Enhancement of the anaerobic hydrolysis and fermentation of municipal solid waste in leachbed reactors by varying flow direction during water addition and leachate recycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uke, Matthew N., E-mail: cnmnu@leeds.ac.uk; Stentiford, Edward

2013-06-15

Highlights: ► Combined downflow and upflow water addition improved hydraulic conductivity. ► Upflow water addition unclogged perforated screen leading to more leachate flow. ► The volume of water added and transmitted positively correlated with hydrolysis process. ► Combined downflow and upflow water addition increased COD production and yield. ► Combined downflow and upflow leachate recycle improved leachate and COD production. - Abstract: Poor performance of leachbed reactors (LBRs) is attributed to channelling, compaction from waste loading, unidirectional water addition and leachate flow causing reduced hydraulic conductivity and leachate flow blockage. Performance enhancement was evaluated in three LBRs M, D andmore » U at 22 ± 3 °C using three water addition and leachate recycle strategies; water addition was downflow in D throughout, intermittently upflow and downflow in M and U with 77% volume downflow in M, 54% volume downflow in U while the rest were upflow. Leachate recycle was downflow in D, alternately downflow and upflow in M and upflow in U. The strategy adopted in U led to more water addition (30.3%), leachate production (33%) and chemical oxygen demand (COD) solubilisation (33%; 1609 g against 1210 g) compared to D (control). The total and volatile solids (TS and VS) reductions were similar but the highest COD yield (g-COD/g-TS and g-COD/g-VS removed) was in U (1.6 and 1.9); the values were 1.33 and 1.57 for M, and 1.18 and 1.41 for D respectively. The strategy adopted in U showed superior performance with more COD and leachate production compared to reactors M and D.« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-06-19

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

PubMed

An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Chan, Y L; Chang, J F; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guo, L; Guo, X H; Guo, Y H; Guo, Z; Hackenburg, R W; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hsiung, Y B; Hu, B Z; Hu, T; Huang, E C; Huang, H X; Huang, X T; Huang, Y B; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jen, K L; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Jones, D; Kang, L; Kettell, S H; Khan, A; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Leung, J K C; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y-C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Mitchell, I; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Qiu, R M; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Stoler, P; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C-H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Yang, Y Z; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, C C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, R; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhou, L; Zhuang, H L; Zou, J H

2017-06-23

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW_{th} reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective ^{239}Pu fission fractions F_{239} from 0.25 to 0.35, Daya Bay measures an average IBD yield σ[over ¯]_{f} of (5.90±0.13)×10^{-43}  cm^{2}/fission and a fuel-dependent variation in the IBD yield, dσ_{f}/dF_{239}, of (-1.86±0.18)×10^{-43}  cm^{2}/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the ^{239}Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes ^{235}U, ^{239}Pu, ^{238}U, and ^{241}Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10^{-43}  cm^{2}/fission have been determined for the two dominant fission parent isotopes ^{235}U and ^{239}Pu. A 7.8% discrepancy between the observed and predicted ^{235}U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Chan, Y. L.; Chang, J. F.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guo, L.; Guo, X. H.; Guo, Y. H.; Guo, Z.; Hackenburg, R. W.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huang, Y. B.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jen, K. L.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Kang, L.; Kettell, S. H.; Khan, A.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Qiu, R. M.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Stoler, P.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Yang, Y. Z.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, C. C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, R.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhou, L.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2017-06-01

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 G Wth reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F239 from 0.25 to 0.35, Daya Bay measures an average IBD yield σ¯f of (5.90 ±0.13 )×10-43 cm2/fission and a fuel-dependent variation in the IBD yield, d σf/d F239, of (-1.86 ±0.18 )×10-43 cm2/fission . This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1 σ . This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17 ±0.17 ) and (4.27 ±0.26 )×10-43 cm2 /fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE Office of Scientific and Technical Information (OSTI.GOV)

An, F. P.; Balantekin, A. B.; Band, H. R.

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

AN EVALUATION OF HEAVY WATER REACTORS FOR POWER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herron, D.P.; Newkirk, W.H.; Puishes, A.

1957-10-01

Reference designs for pressurized and direct-boiling D/sub 2/O reactors were prepared for electrical outputs of 20, 100, and 250 electrical Mw. A number of possible core designs were considered and those utilized which seemed most appropriate to give low-cost power. The technology and costs available today were employed in the preparation of the over-all plant designs. The Consolidated Western Steel Division of U. S. Steel Corporation assisted by preparing a comprehensive report on the design of large pressure vessels and containment vessels. Zr-clad U fuel elements were used as the study basis, but the effect of using UO/sub 2/ andmore » stainless steel cladding was also considered. The principal results found were: (1) Over a wide range of operating conditions snd economic situations, enriched U (up to perhaps 1.4% U/sup 235/) is presently more economic to employ in D/sub 2/O reactors than is natural U. (2) In the longer range, the use of natural U may become more economic as Zr fabrication costs decrease, continuous charge-discharge devices are developed to permit longer exposure levels, and pressure-vessel technology advances so that the large critical masses and core diameters required are not such sn economic penalty on the natural U. The results agree quite well with the data and discussions of the Canadians. (auth)« less

Nuclear reactor for breeding U.sup.233

DOEpatents

Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

1976-01-01

A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

Reactor on-off antineutrino measurement with KamLAND

NASA Astrophysics Data System (ADS)

Gando, A.; Gando, Y.; Hanakago, H.; Ikeda, H.; Inoue, K.; Ishidoshiro, K.; Ishikawa, H.; Koga, M.; Matsuda, R.; Matsuda, S.; Mitsui, T.; Motoki, D.; Nakamura, K.; Obata, A.; Oki, A.; Oki, Y.; Otani, M.; Shimizu, I.; Shirai, J.; Suzuki, A.; Takemoto, Y.; Tamae, K.; Ueshima, K.; Watanabe, H.; Xu, B. D.; Yamada, S.; Yamauchi, Y.; Yoshida, H.; Kozlov, A.; Yoshida, S.; Piepke, A.; Banks, T. I.; Fujikawa, B. K.; Han, K.; O'Donnell, T.; Berger, B. E.; Learned, J. G.; Matsuno, S.; Sakai, M.; Efremenko, Y.; Karwowski, H. J.; Markoff, D. M.; Tornow, W.; Detwiler, J. A.; Enomoto, S.; Decowski, M. P.

2013-08-01

The recent long-term shutdown of Japanese nuclear reactors has resulted in a significantly reduced reactor ν¯e flux at KamLAND. This running condition provides a unique opportunity to confirm and constrain backgrounds for the reactor ν¯e oscillation analysis. The data set also has improved sensitivity for other ν¯e signals, in particular ν¯e’s produced in β-decays from U238 and Th232 within the Earth’s interior, whose energy spectrum overlaps with that of reactor ν¯e’s. Including constraints on θ13 from accelerator and short-baseline reactor neutrino experiments, a combined three-flavor analysis of solar and KamLAND data gives fit values for the oscillation parameters of tan⁡2θ12=0.436-0.025+0.029, Δm212=7.53-0.18+0.18×10-5eV2, and sin⁡2θ13=0.023-0.002+0.002. Assuming a chondritic Th/U mass ratio, we obtain 116-27+28 ν¯e events from U238 and Th232, corresponding to a geo ν¯e flux of 3.4-0.8+0.8×106cm-2s-1 at the KamLAND location. We evaluate various bulk silicate Earth composition models using the observed geo ν¯e rate.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-11-03

separation list: ! 8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction...facilities like reprocessing and enrichment plants and breeder reactors could be viewed as providing a significant nonproliferation benefit because the... breeder reactors would support the 2002 U.S. National Strategy to Combat Weapons of Mass Destruction, in which the United States pledged to “continue to

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-02

8 indigenous Indian power reactors ! Fast Breeder test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment... breeder reactors could be viewed as providing a significant nonproliferation benefit because the materials produced by these plants are a few steps closer...to potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National

FAST NEUTRONIC REACTOR

DOEpatents

Snell, A.H.

1957-12-01

This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

Soluciones analiticas AL problema de jets con velocidad de eyeccion variable EN EL tiempo.

NASA Astrophysics Data System (ADS)

Canto, J.; Raga, A. C.; D'Alessio, P.

1998-11-01

Se presenta un nuevo metodo que permite resolver de manera exacta y analitica las ecuaciones que describen un jet hipersonico con velocidad de eyeccion variable en el tiempo. El metodo se basa en consideraciones sencillas de conservacion de momento para las superficies de trabajo que se forman en el interior del jet. Como ejemplo, se presentan soluciones para jets con variacion sinusoidal en la velocidad de eyeccion, y tambien para el caso de un incremento lineal en el tiempo. Estas soluciones analiticas tienen una clara aplicacion en la interpretacion de las observaciones de jets asociados a objetos Herbig-Haro.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frybort, Jan

A critical experiment is a standard part of training of students at the Training Reactor VR-1 operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. In autumn 2005 the HEU fuel IRT-3M with enrichment 36 % {sup 235}U was replaced by the LEU fuel IRT-4M with enrichment 19.7 % {sup 235}U. The fuel replacement at the VR-1 Reactor is a part of an international program RERTR. This Paper presents basic information about preparation for the fuel replacement and approaching of the first critical state with the new zone configuration C1 which replacedmore » B1 core with the old IRT-3M fuel. The whole process was carried out according to the Czech law and the relevant international recommendations. The experience with the VR-1 operation confirms the assumption that the C1 core configuration will be suitable from the point of view of the reactivity balance for the long term safe operation of the Training Reactor VR-1. (author)« less

Detecting Dark Photons with Reactor Neutrino Experiments.

PubMed

Park, H K

2017-08-25

We propose to search for light U(1) dark photons, A^{'}, produced via kinetically mixing with ordinary photons via the Compton-like process, γe^{-}→A^{'}e^{-}, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ε, the A^{'}-γ mixing parameter, ε, for dark-photon masses below 1 MeV of ε<1.3×10^{-5} and ε<2.1×10^{-5}, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

126. ARAII Plot plan showing location of SL1 power plant ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

126. ARA-II Plot plan showing location of SL-1 power plant (reactor) building, and planned location of administrative and technical support building. C.A. Sundberg and Associates 866-area/ALPR-606-U-1. Date: May 1958. Ineel index code no. 070-0100-00-822-102834. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa

2002-07-01

A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled coremore » has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)« less

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

NASA Astrophysics Data System (ADS)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Method development and validation for simultaneous determination of IEA-R1 reactor’s pool water uranium and silicon content by ICP OES

NASA Astrophysics Data System (ADS)

Ulrich, J. C.; Guilhen, S. N.; Cotrim, M. E. B.; Pires, M. A. F.

2018-03-01

IPEN’s research reactor, IEA-R1, an open pool type research reactor moderated and cooled by light water. High quality water is a key factor in preventing the corrosion of the spent fuel stored in the pool. Leaching of radionuclides from the corroded fuel cladding may be prevented by an efficient water treatment and purification system. However, as a safety management policy, IPEN has adopted a water chemistry control which periodically monitors the levels of uranium (U) and silicon (Si) in the pool’s reactor, since IEA-R1 employs U3Si2-Al dispersion fuel. An analytical method was developed and validated for the determination of uranium and silicon by ICP OES. This work describes the validation process, in a context of quality assurance, including the parameters selectivity, linearity, quantification limit, precision and recovery.

Summary of a joint US-Japan study of potential approaches to reduce the attractiveness of various nuclear materials for use in a nuclear explosive device by a terrorist group

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bathke, C.G.; Inoue, N.; Kuno, Y.

2013-07-01

This paper summarizes the results of a joint US-Japan study to establish a mutual understanding, through scientific-based study, of potential approaches to reduce the attractiveness of various nuclear materials for use in a terrorist nuclear explosive device (NED). 4 approaches that can reduce materials attractiveness with a very high degree of effectiveness are: -) diluting HEU with natural or depleted U to an enrichment of less than 10% U-235; -) storing Pu in nuclear fuel that is not man portable and with a dose rate greater or equal to 10 Gy/h at 1 m; -) storing Pu or HEU inmore » heavy items, i.e. not transportable, provided the removal of the Pu or HEU from the item requires a purification/processing capability; and -) converting Pu and HEU to very dilute forms (such as wastes) that, without any security barriers, would require very long acquisition times to acquire a Category I quantity of Pu or of HEU. 2 approaches that can reduce materials attractiveness with a high degree of effectiveness are: -) converting HEU-fueled research reactors into LEU-fueled research reactors or dilute HEU with natural or depleted U to an enrichment of less than 20% U-235; -) converting U/Al reactor fuel into U/Si reactor fuel. Other approaches have been assessed as moderately or totally inefficient to reduce the attractiveness of nuclear materials.« less

134. ARAII SL1 decontamination and lay down building (ARA614) erected ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

134. ARA-II SL-1 decontamination and lay down building (ARA-614) erected after accidental explosion of SL-1 reactor. Shows vicinity map, index of related drawings, plot plan and other detail. F.C. Torkelson Company 842-area/SL-1-101-U-2. Date: September 1962. Ineel index code no. 070-0101-65-851-150713. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Immobilization of U(VI) from Oxic Groundwater by Hanford 300 Area Sediments and Effects of Columbia River Water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, B.; Cao, Bin; Mishra, Bhoopesh

2012-09-23

Regions within the U.S. Department of Energy Hanford 300 Area (300 A) site experience periodic hydrologic influences from the nearby Columbia River as a result of changing river stage, which causes changes in groundwater elevation, flow direction and water chemistry. An important question is the extent to which the mixing of Columbia River water and groundwater impacts the speciation and mobility of uranium (U). In this study, we designed experiments to mimic interactions among U, oxic groundwater or Columbia River water, and 300 A sediments in the subsurface environment of Hanford 300 A. The goals were to investigate mechanisms of:more » 1) U immobilization in 300 A sediments under bulk oxic conditions and 2) U remobilization from U-immobilized 300 A sediments exposed to oxic Columbia River water. Initially, 300 A sediments in column reactors were fed with U(VI)-containing oxic 1) synthetic groundwater (SGW), 2) organic-amended SGW (OA-SGW), and 3) de-ionized (DI) water to investigate U immobilization processes. After that, the sediments were exposed to oxic Columbia River water for U remobilization studies. The results reveal that U was immobilized by 300 A sediments predominantly through reduction (80-85%) when the column reactor was fed with oxic OA-SGW. However, U was immobilized by 300 A sediments through adsorption (100%) when the column reactors were fed with oxic SGW or DI water. The reduced U in the 300 A sediments fed with OA-SGW was relatively resistant to remobilization by oxic Columbia River water. Oxic Columbia River water resulted in U remobilization (~7%) through desorption, and most of the U that remained in the 300 A sediments fed with OA-SGW (~93%) was in the form of uraninite nanoparticles. These results reveal that: 1) the reductive immobilization of U through OA-SGW stimulation of indigenous 300 A sediment microorganisms may be viable in the relatively oxic Hanford 300 A subsurface environments and 2) with the intrusion of Columbia River water, desorption may be the primary process resulting in U remobilization from OA-SGW-stimulated 300 A sediments at the subsurface of the Hanford 300 A site.« less

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

11. Building Layout, 185189 D, U.S. Atomic Energy Commission, Richland ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

11. Building Layout, 185-189 D, U.S. Atomic Energy Commission, Richland Operations Office, Dwg. No. H-1-14844, 1957. - D-Reactor Complex, Deaeration Plant-Refrigeration Buildings, Area 100-D, Richland, Benton County, WA

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

USGS Publications Warehouse

Nagy, B.; Gauthier-Lafaye, F.; Holliger, P.; Davis, D.W.; Mossman, D.J.; Leventhal, J.S.; Rigali, M.J.; Parnell, J.

1991-01-01

SOME of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter1,2, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid3,4. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite5. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the resolidified, graphitic bitumen. Unlike water-soluble (humic) organic matter, the graphitic bituminous organics at Oklo thus enhanced radionu-clide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lan-thanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pretreated nuclear waste warrants further investigation. ?? 1991 Nature Publishing Group.

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-17

safeguards-irrelevant.” The following facilities and activities were not on the separation list: ! 8 indigenous Indian power reactors ! Fast Breeder ...test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment facilities ! Spent fuel reprocessing facilities (except...potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National Strategy to

Target-fueled nuclear reactor for medical isotope production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

12. General Arrangement Plan, Building 189D, U.S. Atomic Energy Commission, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

12. General Arrangement Plan, Building 189-D, U.S. Atomic Energy Commission, General Electric Company, Dwg. No. H-1-11068, 1958. - D-Reactor Complex, Deaeration Plant-Refrigeration Buildings, Area 100-D, Richland, Benton County, WA

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

The Japanese aerial attack on Hanford Engineer Works

NASA Astrophysics Data System (ADS)

Clark, Charles W.

The day before the Pearl Harbor attack, December 6, 1941, the University of Chicago Metallurgical Laboratory was given four goals: design a plutonium (Pu) bomb; produce Pu by irradiation of uranium (U); extract Pu from the irradiated U; complete this in time to be militarily significant. A year later the first controlled nuclear chain reaction was attained in Chicago Pile 1 (CP-1). In January 1943, Hanford, WA was chosen as the site of the Pu factory. Neutron irradiation of 238U was to be used to make 239Pu. This was done by a larger version of CP-1, Hanford Reactor B, which went critical in September 1944. By July 1945 it had made enough Pu for two bombs: one used at the Trinity test in July; the other at Nagasaki, Japan in August. I focus on an ironic sidelight to this story: disruption of hydroelectric power to Reactor B by a Japanese fire balloon attack on March 10, 1945. This activated the costly coal-fired emergency backup plant to keep the reactor coolant water flowing, thwarting disaster and vindicating the conservative design of Hanford Engineer Works. Management of the Hanford Engineer Works in World War II, H. Thayer (ASCE Press 1996).

In-reactor performance of LWR-type tritium target rods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Da Cruz, D. F.; Rochman, D.; Koning, A. J.

2012-07-01

This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

Experimental investigations of a uranium plasma pertinent to a self-sustaining plasma source

NASA Technical Reports Server (NTRS)

Schneider, R. T.

1971-01-01

The research is pertinent to the realization of a self-sustained fissioning plasma for applications such as nuclear propulsion, closed cycle MHD power generation using a plasma core reactor, and heat engines such as the nuclear piston engine, as well as the direct conversion of fission energy into optical radiation (nuclear pumped lasers). Diagnostic measurement methods and experimental devices simulating plasma core reactor conditions are discussed. Studies on the following topics are considered: (1) ballistic piston compressor (U-235); (2) high pressure uranium plasma (natural uranium); (3) sliding spark discharge (natural uranium); (4) fission fragment interaction (He-3 and U-235); and (5) nuclear pumped lasers (He-3 and U-235).

141. ARAIII Equipment location plan. Includes list of equipment and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

141. ARA-III Equipment location plan. Includes list of equipment and location in reactor, control, and other buildings. Aerojet-general 880-area/GCRE-101-U-1. Date: February 1958. Ineel index code no. 063-0101-65-013-192508. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Safety Issues at the Defense Production Reactors. A Report to the U.S. Department of Energy.

ERIC Educational Resources Information Center

National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

This report provides an assessment of safety management, safety review, and safety methodology employed by the Department of Energy (DOE) and private contractors. Chapter 1, "The DOE Safety Framework," examines safety objectives for production reactors and processes to implement the objectives. Chapter 2, "Technical Issues,"…

13. Architectural First Floor Plan Buildings 185189 D, U.S. Atomic ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. Architectural First Floor Plan Buildings 185-189 D, U.S. Atomic Energy Commission, General Electric Company, Dwg. NO. H-1-11825, 1959. - D-Reactor Complex, Deaeration Plant-Refrigeration Buildings, Area 100-D, Richland, Benton County, WA

Prospects for improved understanding of isotopic reactor antineutrino fluxes

NASA Astrophysics Data System (ADS)

Gebre, Y.; Littlejohn, B. R.; Surukuchi, P. T.

2018-01-01

Predictions of antineutrino fluxes produced by fission isotopes in a nuclear reactor have recently received increased scrutiny due to observed differences in predicted and measured inverse beta decay (IBD) yields, referred to as the "reactor antineutrino flux anomaly." In this paper, global fits are applied to existing IBD yield measurements to produce constraints on antineutrino production by individual plutonium and uranium fission isotopes. We find that fits including measurements from highly U 235 -enriched cores and fits including Daya Bay's new fuel evolution result produce discrepant best-fit IBD yields for U 235 and Pu 239 . This discrepancy can be alleviated in a global analysis of all data sets through simultaneous fitting of Pu 239 , U 235 , and U 238 yields. The measured IBD yield of U 238 in this analysis is (7.02 ±1.65 )×10-43 cm2/fission , nearly two standard deviations below existing predictions. Future hypothetical IBD yield measurements by short-baseline reactor experiments are examined to determine their possible impact on the global understanding of isotopic IBD yields. It is found that future improved short-baseline IBD yield measurements at both high-enriched and low-enriched cores can significantly improve constraints for U 235 , U 238 , and Pu 239 , providing comparable or superior precision to existing conversion- and summation-based antineutrino flux predictions. Systematic and experimental requirements for these future measurements are also investigated.

Studies on magnetocaloric and magnetic coupling effects =

NASA Astrophysics Data System (ADS)

Amaral, Joao Cunha de Sequeira

O presente trabalho apresenta novas metodologias desenvolvidas para a analise das propriedades magneticas e magnetocaloricas de materiais, sustentadas em consideracoes teoricas a partir de modelos, nomeadamente a teoria de transicoes de fase de Landau, o modelo de campo medio molecular e a teoria de fenomeno critico. Sao propostos novos metodos de escala, permitindo a interpretacao de dados de magnetizacao de materiais numa perspectiva de campo medio molecular ou teoria de fenomeno critico. E apresentado um metodo de estimar a magnetizacao espontanea de um material ferromagnetico a partir de relacoes entropia/magnetizacao estabelecidas pelo modelo de campo medio molecular. A termodinamica das transicoes de fase magneticas de primeira ordem e estudada usando a teoria de Landau e de campo medio molecular (modelo de Bean-Rodbell), avaliando os efeitos de fenomenos fora de equilibrio e de condicoes de mistura de fase em estimativas do efeito magnetocalorico a partir de medidas magneticas. Efeitos de desordem, interpretados como uma distribuicao na interaccao magnetica entre ioes, estabelecem os efeitos de distribuicoes quimicas/estruturais nas propriedades magneticas e magnetocaloricas de materiais com transicoes de fase de segunda e de primeira ordem. O uso das metodologias apresentadas na interpretacao das propriedades magneticas de variados materiais ferromagneticos permitiu obter: 1) uma analise quantitativa da variacao de spin por iao Gadolinio devido a transicao estrutural do composto Gd5Si2Ge2, 2) a descricao da configuracao de cluster magnetico de ioes Mn na fase ferromagnetica em manganites da familia La-Sr e La-Ca, 3) a determinacao dos expoentes criticos β e δ do Niquel por metodos de escala, 4) a descricao do efeito da pressao nas propriedades magneticas e magnetocaloricas do composto LaFe11.5Si1.5 atraves do modelo de Bean-Rodbell, 5) uma estimativa da desordem em manganites ferromagneticas com transicoes de segunda e primeira ordem, 6) uma descricao de campo medio das propriedades magneticas da liga Fe23Cu77, 7) o estudo de efeitos de separacao de fase na familia de compostos La0.70-xErxSr0.30MnO3 e 8) a determinacao realista da variacao de entropia magnetica na familia de compostos de efeito magnetocalorico colossal Mn1-x-yFexCryAs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

Rotating Fluidized Bed Reactor for Space Nuclear Propulsion. Annual Report; Design Studies and Experimental Results

NASA Technical Reports Server (NTRS)

1971-01-01

The rotating fluidized bed reactor concept is being investigated for possible application in nuclear propulsion systems. Physics calculations show U-233 to be superior to U-235 as a fuel for a cavity reactor of this type. Preliminary estimates of the effect of hydrogen in the reactor, reflector material, and power peaking are given. A preliminary engineering analysis was made for U-235 and U-233 fueled systems. An evaluation of the parameters affecting the design of the system is given, along with the thrust-to-weight ratios. The experimental equipment is described, as are the special photographic techniques and procedures. Characteristics of the fluidized bed and experimental results are given, including photographic evidence of bed fluidization at high rotational velocities.

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

NASA Astrophysics Data System (ADS)

Ioffe, B. L.; Kochurov, B. P.

2012-02-01

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

Nuclear Propulsion for Space Applications

NASA Technical Reports Server (NTRS)

Houts, M. G.; Bechtel, R. D.; Borowski, S. K.; George, J. A.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Adams, R. B.

2013-01-01

Basics of Nuclear Systems: Long history of use on Apollo and space science missions. 44 RTGs and hundreds of RHUs launched by U.S. during past 4 decades. Heat produced from natural alpha (a) particle decay of Plutonium (Pu-238). Used for both thermal management and electricity production. Used terrestrially for over 65 years. Fissioning 1 kg of uranium yields as much energy as burning 2,700,000 kg of coal. One US space reactor (SNAP-10A) flown (1965). Former U.S.S.R. flew 33 space reactors. Heat produced from neutron-induced splitting of a nucleus (e.g. U-235). At steady-state, 1 of the 2 to 3 neutrons released in the reaction causes a subsequent fission in a "chain reaction" process. Heat converted to electricity, or used directly to heat a propellant. Fission is highly versatile with many applications.

Oxygen potentials of mixed oxide fuels for fast reactors

NASA Astrophysics Data System (ADS)

Kato, M.; Tamura, T.; Konashi, K.

2009-03-01

Oxygen potentials of homogenous (Pu0.2U0.8)O2-x and (Am0.02Pu0.30Np0.02U0.66)O2-x which have been developed as fuels for fast breeder reactors were measured at temperatures of 1473-1623 K by a gas equilibrium method using an (Ar, H2, H2O) gas mixture. The measured oxygen potentials of (Pu0.2U0.8)O2-x were about 25 kJ mol-1 lower than those of (Pu0.3U0.7)O2-x measured previously and were consistent with the values calculated by Besmann and Lindemer's model. The measured oxygen potentials of (Am0.02Pu0.30Np0.02U0.66)O2-x were slightly higher than those of MOX without minor actinides. No fuel-cladding chemical interaction is affected significantly by adding their minor actinides.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benton, J; Wall, D; Parker, E

This paper presents the latest information on one of the Accelerated Highly Enriched Uranium (HEU) Disposition initiatives that resulted from the May 2002 Summit meeting between Presidents George W. Bush and Vladimir V. Putin. These initiatives are meant to strengthen nuclear nonproliferation objectives by accelerating the disposition of nuclear weapons-useable materials. The HEU Transparency Implementation Program (TIP), within the National Nuclear Security Administration (NNSA) is working to implement one of the selected initiatives that would purchase excess Russian HEU (93% 235U) for use as fuel in U.S. research reactors over the next ten years. This will parallel efforts to convertmore » the reactors' fuel core from HEU to low enriched uranium (LEU) material, where feasible. The paper will examine important aspects associated with the U.S. research reactor HEU purchase. In particular: (1) the establishment of specifications for the Russian HEU, and (2) transportation safeguard considerations for moving the HEU from the Mayak Production Facility in Ozersk, Russia, to the Y-12 National Security Complex in Oak Ridge, TN.« less

Immobilization of U(VI) from oxic groundwater by Hanford 300 Area sediments and effects of Columbia River water.

PubMed

Ahmed, Bulbul; Cao, Bin; Mishra, Bhoopesh; Boyanov, Maxim I; Kemner, Kenneth M; Fredrickson, Jim K; Beyenal, Haluk

2012-09-01

Regions within the U.S. Department of Energy Hanford 300 Area (300 A) site experience periodic hydrologic influences from the nearby Columbia River as a result of changing river stage, which causes changes in groundwater elevation, flow direction and water chemistry. An important question is the extent to which the mixing of Columbia River water and groundwater impacts the speciation and mobility of uranium (U). In this study, we designed experiments to mimic interactions among U, oxic groundwater or Columbia River water, and 300 A sediments in the subsurface environment of Hanford 300 A. The goals were to investigate mechanisms of: 1) U immobilization in 300 A sediments under bulk oxic conditions and 2) U remobilization from U-immobilized 300 A sediments exposed to oxic Columbia River water. Initially, 300 A sediments in column reactors were fed with U(VI)-containing oxic 1) synthetic groundwater (SGW), 2) organic-amended SGW (OA-SGW), and 3) de-ionized (DI) water to investigate U immobilization processes. After that, the sediments were exposed to oxic Columbia River water for U remobilization studies. The results reveal that U was immobilized by 300 A sediments predominantly through reduction (80-85%) when the column reactor was fed with oxic OA-SGW. However, U was immobilized by 300 A sediments through adsorption (100%) when the column reactors were fed with oxic SGW or DI water. The reduced U in the 300 A sediments fed with OA-SGW was relatively resistant to remobilization by oxic Columbia River water. Oxic Columbia River water resulted in U remobilization (∼7%) through desorption, and most of the U that remained in the 300 A sediments fed with OA-SGW (∼93%) was in the form of uraninite nanoparticles. These results reveal that: 1) the reductive immobilization of U through OA-SGW stimulation of indigenous 300 A sediment microorganisms may be viable in the relatively oxic Hanford 300 A subsurface environments and 2) with the intrusion of Columbia River water, desorption may be the primary process resulting in U remobilization from OA-SGW-stimulated 300 A sediments at the subsurface of the Hanford 300 A site. Copyright © 2012 Elsevier Ltd. All rights reserved.

Dynamics of heat-pipe reactors

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1971-01-01

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

METALLURGY DIVISION QUARTERLY REPORT FOR JULY, AUGUST, AND SEPTEMBER 1957

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1958-10-01

Advanced Water Reactor Program. Three firings were made of initial closed-porosity fuel pellet bodies. Each firing coatained pellets of the composition 90 wt.% ThO/sub 2/-10 wt.%fl U0/sub 2/ with various additives and firing variables. Fast Power Breeder Reactor Program. To determine the potential usefulness of a Zr-5 wt. % Pu alloy, the fabricability of the alloy was tested. The manufacture of rod stock from which fuel and blanket elements for the Mark III loading of the EBR-1 were prcduced has been completed. The effect of irradiation on extruded and heat-treated U-2 wt.% Zr alloy for the EBR- 1 is reported.more » Fabrication procedures for making graphite-U/sub 3/O/sub 8/ test specimens for the TREAT Reactor were investigated. Advanced Engineering and Development. Ultrasonic bond tests were conducted on 590 EBR-1 Mark III blanket fuel elemeats. The blanket rods and part of the fuel rcds for the EBR-1 Mark III loading are being checked for cladding thickness by an eddy current system. Investigations of corrosionresistant Zr-Nb alloy were coatinued. Corrosion of MR alloys is being studied Ln support of the Mighty Mouse reactor program. Dynamic corrosion tests were performed on aluminum alloys, and results are included. Prcduction, Treatment, and Properties of Materials. The progress of the program of preparing highpurity Pu by fused salt electrolysis is summarized. Velocities of ultrasonic waves propagated in directions suitable for determining the room- temperature elastic moduli C/sub 12/, C/sub 13/, and C/sub 23/ of alpha U were determined. investigation of recrystallization in heavily coldrolled alpha- uranium sheet without a texture change was essentially concluded during this quarter. Selfdiffasion runs in polycrystalline uranium in the gamma phase, using the sputtering technique, have yielded a tentative value for the diffusion coefficient between 10/sup -8/ and 10/sup -7/ cm/sup 2/second. The preparation of high-purity U-Pan alloys is reponted. The data for the alpha-tobeta transformation temperatures in high-purity U and U-C alloys were confirmed by experiments on new specimens. Microstructure, density, and thermal arrest data were obtained for several injection cast, nominal U-5 wt.%fl fissium and U-8 wt.%fl fissium alloys. Phase diagrams are preseated for U-Mo and U-Ru alloys. Alloy Theory and The Nature of Solids. Four new isomorphs of Ti/sub 2/Ni have been discovered. Effects of Irradiation on Materials. The experimental and analytical work on the radial distribution of thermal neutrons within cylindrically shaped fuel specimens during irradiation was completed. (For preceding period see ANL-5790.) (W.L.H.)« less

Assessment of nuclear reactor concepts for low power space applications

NASA Technical Reports Server (NTRS)

Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

1988-01-01

The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaffke, Patrick John

This acts as a short note on the effects of varying the value of the endpoints of the thermal, epithermal, and fast flux groups. As expected, varying these endpoints can alter the value of the cross-section for a given nuclide. This effect is quantified in this note for an important nuclide in reactor simulations, 238U. Uranium-238 is responsible for the production of Plutonium in most reactors, making it critical to understand all of the 238U capture modes leading to Plutonium. We explicitly quantify the reaction rates for 238U that are altered when we use a given research reactor fluxmore » and vary the endpoint definitions of said flux as well as the reactor position.« less

The basic features of a closed fuel cycle without fast reactors

NASA Astrophysics Data System (ADS)

Bobrov, E. A.; Alekseev, P. N.; Teplov, P. S.

2017-01-01

In this paper the basic features of a closed fuel cycle with thermal reactors are considered. The three variants of multiple Pu and U recycling in VVER reactors was investigated. The comparison of MOX and REMIX fuel approaches for closed fuel cycle with thermal reactors is presented. All variants make possible to recycle several times the total amount of Pu and U obtained from spent fuel. The reported study was funded by RFBR according to the research project № 16-38-00021

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

Antineutrino monitoring of thorium reactors

DOE PAGES

Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

2016-09-30

Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg ofmore » 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. In conclusion, a rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.« less

Antineutrino monitoring of thorium reactors

NASA Astrophysics Data System (ADS)

Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

2016-09-01

Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuel types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring. Thorium molten salt reactors breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg of 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. A rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

Status report on the fusion breeder

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moir, R.W.

1980-12-12

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW m/sup -2/, and the hybrid should cost lessmore » than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are unusually rapid.« less

Measurement of 235U(n,n'γ) and 235U(n,2nγ) reaction cross sections

NASA Astrophysics Data System (ADS)

Kerveno, M.; Thiry, J. C.; Bacquias, A.; Borcea, C.; Dessagne, P.; Drohé, J. C.; Goriely, S.; Hilaire, S.; Jericha, E.; Karam, H.; Negret, A.; Pavlik, A.; Plompen, A. J. M.; Romain, P.; Rouki, C.; Rudolf, G.; Stanoiu, M.

2013-02-01

The design of generation IV nuclear reactors and the studies of new fuel cycles require knowledge of the cross sections of various nuclear reactions. Our research is focused on (n,xnγ) reactions occurring in these new reactors. The aim is to measure unknown cross sections and to reduce the uncertainty on present data for reactions and isotopes of interest for transmutation or advanced reactors. The present work studies the 235U(n,n'γ) and 235U(n,2nγ) reactions in the fast neutron energy domain (up to 20 MeV). The experiments were performed with the Geel electron linear accelerator GELINA, which delivers a pulsed white neutron beam. The time characteristics enable measuring neutron energies with the time-of-flight (TOF) technique. The neutron induced reactions [in this case inelastic scattering and (n,2n) reactions] are identified by on-line prompt γ spectroscopy with an experimental setup including four high-purity germanium (HPGe) detectors. A fission ionization chamber is used to monitor the incident neutron flux. The experimental setup and analysis methods are presented and the model calculations performed with the TALYS-1.2 code are discussed.

The chemical state of defective uranium-plutonium oxide fuel pins irradiated in sodium cooled reactors

NASA Astrophysics Data System (ADS)

Kleykamp, H.

1997-09-01

Steady-state irradiation experiments were conducted in the sodium loop of the Siloe reactor on artificially failed mixed oxide pins that had been pre-irradiated in fast reactors up to 11.5% burnup. The formation of the predominant reaction product Na 3(U,Pu)O 4 starts on the fuel surface and is terminated when a lower O/(U + Pu) threshold of the fuel is attained. The axial extent of the reaction product depends on the size of the initial cladding defect. The occurrence of secondary cracks is possible. Na(U,Pu)O 3 forms at higher fuel temperatures. The existence of Na 3U 1- xPu xO 4 is shown in pre-irradiated blanket pins after artificial defect formation. Caesium in the oxocompounds is reduced to the metallic state and is dissolved in the coolant. Evidence of a very low chemical potential of oxygen in defective fuel pins is sustained by the occurrence of actinide-platinum metal phases formed by coupled reduction of hypostoichiometric fuel with ɛ-(Mo,Tc,Ru,Rh,Pd) precipitates. Continued operation of defective pins is not hazardous by easy precautions.

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

U-PuO2, U-PuC, U-PuN cermet fuel for fast reactor

NASA Astrophysics Data System (ADS)

Mishra, Sudhir; Kaity, Santu; Banerjee, Joydipta; Nandi, Chiranjeet; Dey, G. K.; Khan, K. B.

2018-02-01

Cermet fuel combines beneficial properties of both ceramic and metal and attracts global interest for research as a candidate fuel for nuclear reactors. In the present study, U matrix PuC/PuN/PuO2 cermet for fast reactor have been fabricated on laboratory scale by the powder metallurgy route. Characterization of the fuel has been carried out using Dilatometer, Differential Thermal analysis (DTA), X-ray diffractometer and Optical microscope. X ray diffraction study of the fuel reveals presence of different phases. The PuN dispersed cermet was observed to have high solidus temperature as compared to PuC and PuO2 dispersed cermet. Swelling was observed in U matrix PuO2 cermet which also showed higher thermal expansion. Among the three cermets studied, U matrix PuC cermet showed maximum thermal conductivity.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Richards, Matt; Hamilton, Chris

This report provides supplemental information to the assessment of target markets provided in Appendix A of the 2012 Next Generation Nuclear Plant (NGNP) Industry Alliance (NIA) business plan [NIA 2012] for deployment of High Temperature Gas-Cooled Reactors (HTGRs) in the 2025 – 2050 time frame. This report largely reiterates the [NIA 2012] assessment for potential deployment of 400 to 800 HTGR modules (100 to 200 HTGR plants with 4 reactor modules) in the 600-MWt class in North America by 2050 for electricity generation, co-generation of steam and electricity, oil sands operations, hydrogen production, and synthetic fuels production (e.g., coal tomore » liquids). As the result of increased natural gas supply from hydraulic fracturing, the current and historically low prices of natural gas remain a significant barrier to deployment of HTGRs and other nuclear reactor concepts in the U.S. However, based on U.S. Department of Energy (DOE) Energy Information Agency (EIA) data, U.S. natural gas prices are expected to increase by the 2030 – 2040 timeframe when a significant number of HTGR modules could be deployed. An evaluation of more recent EIA 2013 data confirms the assumptions in [NIA 2012] of future natural gas prices in the range of approximately $7/MMBtu to $10/MMBtu during the 2030 – 2040 timeframe. Natural gas prices in this range will make HTGR energy prices competitive with natural gas, even in the absence of carbon-emissions penalties. Exhibit ES-1 presents the North American projections in each market segment including a characterization of the market penetration logic. Adjustments made to the 2012 data (and reflected in Exhibit ES-1) include normalization to the slightly larger 625MWt reactor module, segregation between steam cycle and more advanced (higher outlet temperature) modules, and characterization of U.S. synthetic fuel process applications as a separate market segment.« less

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-07-17

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...planned nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008...contracting between U.S. firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power

Burning high-level TRU waste in fusion fission reactors

NASA Astrophysics Data System (ADS)

Shen, Yaosong

2016-09-01

Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Computational Fluid Dynamics.

DTIC Science & Technology

1986-06-01

this area concentrated first on the compu- tation of foldpoints by means of one of the many possible forms of augmented equations. One approach in this...this area is, of course, the deter- mination of the principal characteristic features of the solution manifold of the given parametrized equation. This...into Spanish and appeared in Metodos Numericos Para Calculo y Diseno en Ingeniera 1 (1985), 67-80. [5] J. P. Fink and W. C. Rheinboldt, "Folds on the

Distribution of the Atomic Component in the Local Interstellar Medium. the Warm Neutral Intercloud Medium (wnm).

NASA Astrophysics Data System (ADS)

Poppel, W. G. L.; Marronetti, P.; Benagalia, P.

1990-11-01

RESUMEN : Se do'-- ' ti--'s `3' pCrfii ":-::' in:- a JI- nea de 2J rrn dci HI para Cs r%ud. r i com -." ..nte bia 'a del medjo intere=-telar. l metodo ce baa n ci :..: i3:'-i oaussiano de las contr'ibucioncc- de di-ha corno': ..:c'n+. a en los oerfiles. Los r'inc oaic :. r..' : t":. o r.-"r:st. n resLtfnidci.--' en las Fipuras 1 2 u.''e dan J `:- re ' ec:t.i' e"'.:.-. jl'-.'.. -. tribucic-es de la densidad dE? columr N)H `/ .ta \\` .`i(:) ,fj,:,'t.' radial V en funri6n de I. a osic:i6n 1., : Two atlas of orofiles of .1-cf; ha'.' ' been tr,'E-:"::i for computino the HI--emission d'. e to Lhe j (He i :-:: ` H -).- bina 1974 and Colomb,, Po"ppel and Hei le . .4'..) . ih& m: thod is based on a #` anpr'o>'imation Cf the rontr-ibutior:s of the WNM to the profiles. The main are summ-. 'i:.ed here in the form of two maps .howin the di :'.tributions of and of the radial velocit respectivel as a ..r,t.ion of and b (Fios. I and 2). Key : INTERSTELLAR-CLOUDS - INTERSTELLAR-GENERAL - GALAXY- STRUCTURE

Measurement of the^ 235U(n,n')^235mU Integral Cross Section in a Pulsed Reactor

NASA Astrophysics Data System (ADS)

Vieira, D. J.; Bond, E. M.; Belier, G.; Meot, V.; Becker, J. A.; Macri, R. A.; Authier, N.; Hyneck, D.; Jacquet, X.; Jansen, Y.; Legrendre, J.

2009-10-01

We will present the integral measurement of the neutron inelastic cross section of ^235U leading to the 26-minute, E*=76.5 eV isomer state. Small samples (5-20 microgm) of isotope-enriched ^235U were activated in the central cavity of the CALIBAN pulsed reactor at Valduc where a nearly pure fission neutron spectrum is produced with a typical fluence of 3x10^14 n/cm^2. After 30 minutes the samples were removed from the reactor and counted in an electrostatic-deflecting electron spectrometer that was optimized for the detection of ^235mU conversion electrons. From the decay curve analysis of the data, the 26-minute ^235mU component was extracted. Preliminary results will be given and compared to gamma-cascade calculations assuming complete K-mixing or with no K-mixing.

Anaerobic digestion of macroalgae: methane potentials, pre-treatment, inhibition and co-digestion.

PubMed

Nielsen, H B; Heiske, S

2011-01-01

In the present study we tested four macroalgae species--harvested in Denmark--for their suitability of bioconversion to methane. In batch experiments (53 degrees C) methane yields varied from 132 ml g volatile solids(-1) (VS) for Gracillaria vermiculophylla, 152 mi gVS(-1) for Ulva lactuca, 166 ml g VS(-1) for Chaetomorpha linum and 340 ml g VS(-1) for Saccharina latissima following 34 days of incubation. With an organic content of 21.1% (1.5-2.8 times higher than the other algae) S. latissima seems very suitable for anaerobic digestion. However, the methane yields of U. lactuca, G. vermiculophylla and C. linum could be increased with 68%, 11% and 17%, respectively, by pretreatment with maceration. U. lactuca is often observed during 'green tides' in Europe and has a high cultivation potential at Nordic conditions. Therefore, U. lactuca was selected for further investigation and co-digested with cattle manure in a lab-scale continuously stirred tank reactor. A 48% increase in methane production rate of the reactor was observed when the concentration of U. lactuca in the feedstock was 40% (VS basis). Increasing the concentration to 50% had no further effect on the methane production, which limits the application of this algae at Danish centralized biogas plant.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, Mitchell T.

Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data.more » Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities. Highlights in this FY2017 report include new insights with respect to the forces required to produce the observed Daiichi Unit 1 (1F1) shield plug endstate, the observed leakage from 1F1 components, and the amount of combustible gas generation required to produce the observed explosions in Daiichi Units 3 and 4 (1F3 and 1F4). This report contains an appendix with a list of examination needs that was updated after U.S. experts reviewed recently obtained information from examinations at Daiichi. Additional details for higher priority, near-term, examination activities are also provided. This report also includes an appendix with a description of an updated website that has been reformatted to better assist U.S. experts by providing information in an archived retrievable location, as well as an appendix summarizing U.S. Forensics activities to host the TMI-2 Knowledge Transfer and Relevance to Fukushima Meeting that was held in Idaho Falls, ID, on October 10-14, 2016.« less

Measurement of the Infinite Multiplication Constant of Natural Uranium--Graphite Lattices in the RB-1 Critical Assembly by Means of the Zero Reactivity Method. MISURA DELLA COSTANTE DI MOLTIPLICAZIONE INFINITA DI RETICOLI A URANIO NATURALE E GRAFITE NELL'INSIEME CRITICO RB-1 CON IL METODO DELLA REATTIVITA' NULLA (in Italian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghillardotti, G.

1966-07-01

To reduce uncertainties to the minimum, measurements in the RB-1 were conducted on the same materials and with the same instrumentation as those used previously in MARIUS. The values measured in the RB-1, compared with the already known substitution data, are as follows: (a) the difference between the multiplication and the absorption intensity; (b) the fine structure of the flux in the cell; (c) the Pu/U index. The infinite mutiplication factor K{sub infinity} is obtained by combining measurements (a) and (b). The results of this research can be summed up as follows: 1. A consistent and complete experimental procedure hasmore » been devised for measuring the K{sub infinity} of natural uranium/graphite lattices by means of the zero reactivity method. The same applies to the procedure for analysis of the experimental data. 2. The error in (K{sub infinity} -- 1) inherent in the measurement can in our opinion be reduced to 2%. This limit was reached in the last experiment on lattices consisting of tubular elements. 3. Agreement proved to be good with the results obtained by the CEA in the critical assembly MARIUS. (auth)« less

Current status of the development of high density LEU fuel for Russian research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

DOE PAGES

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

2015-09-03

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

Production of Thorium-229 at the ORNL High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boll, Rose Ann; Garland, Marc A; Mirzadeh, Saed

The investigation of targeted cancer therapy using -emitters has developed considerably in recent years and clinical trials have generated promising results. In particular, the initial clinical trials for treatment of acute myeloid leukemia have demonstrated the effectiveness of the -emitter 213Bi in killing cancer cells [1]. Pre-clinical studies have also shown the potential application of both 213Bi and its 225Ac parent radionuclide in a variety of cancer systems and targeted radiotherapy [2]. Bismuth-213 is obtained from a radionuclide generator system from decay of the 10-d 225Ac parent, a member of the 7340-y 229Th chain. Currently, 233U is the only viablemore » source for high purity 229Th; however, due to increasing difficulties associated with 233U safeguards, processing additional 233U is presently unfeasible. The recent decision to downblend and dispose of enriched 233U further diminished the prospects for extracting 229Th from 233U stock. Nevertheless, the anticipated growth in demand for 225Ac may soon exceed the levels of 229Th (~40 g or ~8 Ci; ~80 times the current ORNL 229Th stock) present in the aged 233U stockpile. The alternative routes for the production of 229Th, 225Ra and 225Ac include both reactor and accelerator approaches [3]. Here, we describe production of 229Th via neutron transmutation of 226Ra targets in the ORNL High Flux Isotope Reactor (HFIR).« less

Study on underclad cracking in nuclear reactor vessel steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Horiya, T.; Takeda, T.; Yamato, K.

1985-02-01

Susceptibility to underclad cracking in nuclear reactor vessel steels, such as SA533 Grade B Class 1 and SA508 Class 2, was studied in detail. A convenient simulation test method using simulated HAZ specimens of small size has been developed for quantitative evaluation of susceptibility to underclad cracks. The method can predict precisely the cracking behavior in weldments of steels with relative low crack susceptibility. The effect of chemical compositions on susceptibility to the cracking was examined systematically using the developed simulation test method and the following index was obtained from the test results: U = 20(V) + 7(C) + 4(Mo)more » + (Cr) + (Cu) - 0.5(Mn) + 1.5 log(X) X = Al . . . Al/2N less than or equal to 1 X = 2N . . . Al/2N > 1 It was confirmed that the new index (U) is useful for the prediction of crack susceptibility of the nuclear vessel steels; i.e., no crack initiation is detected in weldments in the roller bend test for steels having U value below 0.90.« less

Small low mass advanced PBR's for propulsion

NASA Astrophysics Data System (ADS)

Powell, J. R.; Todosow, M.; Ludewig, H.

1993-10-01

The advanced Particle Bed Reactor (PBR) to be described in this paper is characterized by relatively low power, and low cost, while still maintaining competition values for thrust/weight, specific impulse and operating times. In order to retain competitive values for the thrust/weight ratio while reducing the reactor size, it is necessary to change the basic reactor layout, by incorporating new concepts. The new reactor design concept is termed SIRIUS (Small Lightweight Reactor Integral Propulsion System). The following modifications are proposed for the reactor design to be discussed in this paper: Pre-heater (U-235 included in Moderator); Hy-C (Hydride/De-hydride for Reactor Control); Afterburner (U-235 impregnated into Hot Frit); and Hy-S (Hydride Spike Inside Hot Frit). Each of the modifications will be briefly discussed below, with benefits, technical issues, design approach, and risk levels addressed. The paper discusses conceptual assumptions, feasibility analysis, mass estimates, and information needs.

Innovative approach for benzene degradation using hybrid surface/packed-bed discharge plasmas.

PubMed

Jiang, Nan; Lu, Na; Shang, Kefeng; Li, Jie; Wu, Yan

2013-09-03

An innovative plasma reactor, which generates hybrid surface/packed-bed discharge (HSPBD) plasmas, was employed for the degradation of benzene. The HSPBD reactor was found to display remarkably better benzene degradation, mineralization, and energy performance than surface or packed-bed discharge reactors alone. The degradation efficiency, CO2 selectivity, and energy yield in the HSPBD reactor were 21%, 11%, and 3.9 g kWh-1 higher, respectively, than in a surface discharge reactor and 30%, 21%, and 5.5 g kWh-1 higher, respectively, than in a packed-bed discharge reactor operated at 280 J L-1. Particularly, the benzene degradation in the HSPBD reactor exhibited an unambiguous synergistic enhancement rather than a simple additive effect using the surface discharge and packed-bed discharge reactors. Moreover, in the HSPBD reactor, the formation of byproducts, such as NO2, was suppressed, while O3 was promoted. The use of N2 as the carrier gas was found to be effective for benzene degradation because of the fast reaction rate of N2(A3∑u+) with benzene, and oxygen species derived from the dissociation of O2 were found to be significant in the mineralization process. Thus, the addition of O2 to N2 allows for efficient degradation of benzene, and the optimized amount of O2 was determined to be 3%.

Reactor Neutronics: Impact of Fissile Material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, F.; Hill, R. N.

Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratiomore » between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.« less

Reactor Neutronics: Impact of Fissile Material

DOE PAGES

Heidet, F.; Hill, R. N.

2017-06-09

Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratiomore » between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.« less

Updated global 3+1 analysis of short-baseline neutrino oscillations

NASA Astrophysics Data System (ADS)

Gariazzo, S.; Giunti, C.; Laveder, M.; Li, Y. F.

2017-06-01

We present the results of an updated fit of short-baseline neutrino oscillation data in the framework of 3+1 active-sterile neutrino mixing. We first consider ν e and {\\overline{ν}}_e disappearance in the light of the Gallium and reactor anomalies. We discuss the implications of the recent measurement of the reactor {\\overline{ν}}_e spectrum in the NEOS experiment, which shifts the allowed regions of the parameter space towards smaller values of | U e4|2. The β-decay constraints of the Mainz and Troitsk experiments allow us to limit the oscillation length between about 2 cm and 7 m at 3 σ for neutrinos with an energy of 1 MeV. The corresponding oscillations can be discovered in a model-independent way in ongoing reactor and source experiments by measuring ν e and {\\overline{ν}}_e disappearance as a function of distance. We then consider the global fit of the data on short-baseline {}_{ν_{μ}}^{(-)}{\\to}_{ν_e}^{(-)} transitions in the light of the LSND anomaly, taking into account the constraints from {}_{ν_e}^{(-)} and {}_{ν_{μ}}^{(-)} disappearance experiments, including the recent data of the MINOS and IceCube experiments. The combination of the NEOS constraints on | U e4|2 and the MINOS and IceCube constraints on | U μ4|2 lead to an unacceptable appearance-disappearance tension which becomes tolerable only in a pragmatic fit which neglects the MiniBooNE low-energy anomaly. The minimization of the global χ 2 in the space of the four mixing parameters Δ m 41 2 , | U e4|2, | U μ4|2, and | U τ4|2 leads to three allowed regions with narrow Δ m 41 2 widths at Δ m 41 2 ≈ 1.7 (best-fit), 1.3 (at 2 σ), 2.4 (at 3 σ) eV2. The effective amplitude of short-baseline {}_{ν_{μ}}^{(-)}{\\to}_{ν_e}^{(-)} oscillations is limited by 0.00048 ≲ sin2 2 ϑ eμ ≲ 0.0020 at 3 σ. The restrictions of the allowed regions of the mixing parameters with respect to our previous global fits are mainly due to the NEOS constraints. We present a comparison of the allowed regions of the mixing parameters with the sensitivities of ongoing experiments, which show that it is likely that these experiments will determine in a definitive way if the reactor, Gallium and LSND anomalies are due to active-sterile neutrino oscillations or not.

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Comparison of heavy metal toxicity in continuous flow and batch reactors

NASA Astrophysics Data System (ADS)

Sengor, S. S.; Gikas, P.; Moberly, J. G.; Peyton, B. M.; Ginn, T. R.

2009-12-01

The presence of heavy metals may significantly affect microbial growth. In many cases, small amounts of particular heavy metals may stimulate microbial growth; however, larger quantities may result in microbial growth reduction. Environmental parameters, such as growth pattern may alter the critical heavy metal concentration, above which microbial growth stimulation turns to growth inhibition. Thus, it is important to quantify the effects of heavy metals on microbial activity for understanding natural or manmade biological reactors, either in situ or ex situ. Here we compare the toxicity of Zn and Cu on Arthrobacter sp., a heavy metal tolerant microorganism, under continuous flow versus batch reactor operations. Batch and continuous growth tests of Arthrobacter sp. were carried out at various individual and combined concentrations of Zn and Cu. Biomass concentration (OD) was measured for both the batch and continuous reactors, whereas ATP, oxygen uptake rates and substrate concentrations were additionally measured for the continuous system. Results indicated that Cu was more toxic than Zn under all conditions for both systems. In batch reactors, all tested Zn concentrations up to 150 uM showed a stimulatory effect on microbial growth. However, in the case of mixed Zn and Cu exposures, the presence of Zn either eliminated (at the 50 uM level both Zn and Cu) or reduced by ~25% (at the 100 and 150 uM levels both Zn and Cu) the Cu-induced inhibition. In the continuous system, only one test involved combined Cu (40uM) and Zn (125uM) and this test showed similar results to the 40uM Cu continuous test, i.e., no reduction in inhibition. The specific ATP concentration, i.e., ATP/OD, results for the continuous reactor showed an apparent recovery for both Cu-treated populations, although neither the OD nor glucose data showed any recovery. This may reflect that the individual microorganisms that survived after the addition of heavy metals, kept maintaining the usual ATP levels, as before metal addition. The last may imply a short of adaptation by some microorganisms to the presence of heavy metals. Overall, the batch reactor tests underestimated significantly the heavy metal inhibition, as compared to the continuous flow reactors. Therefore, the results of batch reactor tests should be used with some caution when heavy metal inhibition is to be interpreted for continuous flow natural environmental systems, such as rivers or wetlands.

76 FR 64126 - Advisory Committee on Reactor Safeguards; Procedures for Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Procedures for Meetings.... Nuclear Regulatory Commission's (NRC's) Advisory Committee on Reactor Safeguards (ACRS) pursuant to the... specified in the Federal Register Notice, care of the Advisory Committee on Reactor Safeguards, U.S. Nuclear...

78 FR 35056 - Effectiveness of the Reactor Oversight Process Baseline Inspection Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-11

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0125] Effectiveness of the Reactor Oversight Process... the effectiveness of the reactor oversight process (ROP) baseline inspection program with members of... Nuclear Reactor Regulations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301...

78 FR 67205 - Advisory Committee on Reactor Safeguards; Procedures for Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Procedures for Meetings.... Nuclear Regulatory Commission's (NRC's) Advisory Committee on Reactor Safeguards (ACRS) pursuant to the... specified in the Federal Register Notice, care of the Advisory Committee on Reactor Safeguards, U.S. Nuclear...

S/sub n/ analysis of the TRX metal lattices with ENDF/B version III data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, F.J.; Pearlstein, S.

1975-03-01

Two critical assemblies, designated as thermal-reactor benchmarks TRX-1 and TRX-2 for ENDF/B data testing, were analyzed using the one-dimensional S/sub n/-theory code SCAMP. The two assemblies were simple lattices of aluminum-clad, uranium-metal fuel rods in triangular arrays with D$sub 2$O as moderator and reflector. The fuel was low-enriched (1.3 percent $sup 235$U), 0.387-inch in diameter and had an active height of 48 inches. The volume ratio of water to uranium was 2.35 for the TRX-1 lattice and 4.02 for TRX-2. Full-core S/sub n/ calculations based on Version III data were performed for these assemblies and the results obtained were comparedmore » with the measured values of the multiplication factors, the ratio of epithermal-to-thermal neutron capture in $sup 238$U, the ratio of epithermal-to-thermal fission in $sup 235$U, the ratio of $sup 238$U fission to $sup 235$U fission, and the ratio of capture in $sup 238$U to fission in $sup 235$U. Reaction rates were obtained from a central region of the full- core problems. Multigroup cross sections for the reactor calculation were obtained from S/sub n/ cell calculations with resonance self-shielding calculated using the RABBLE treatment. The results of the analyses are generally consistent with results obtained by other investigators. (auth)« less

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

DTIC Science & Technology

2009-07-01

inalienable right and, by and large, neither have U.S. government officials. However, the case of Iran raises perhaps the most critical question in this...slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of uranium U3O8 to produce 1 lb of low...thermal neutrons but can induce fission in all actinides , including all plutonium isotopes. Therefore, nuclear fuel for a fast reactor must have a

Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

C. Fiorina; N. E. Stauff; F. Franceschini

2013-12-01

The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associatedmore » with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.« less

Status of French reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballagny, A.

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (exceptmore » if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.« less

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-04-08

Kazakhstan might start uranium exports to India in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor...reactors producing more than 5 MW thermal or special nuclear material connected therewith. U.S. Nuclear Cooperation with India: Issues for Congress

Deployment history and design considerations for space reactor power systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2009-05-01

The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

Galerkin's Method and the Double P$sub 1$ approximation for Thermal Flux Calculation; IL METODO DI GALERKIN E LA DOPPIA P$sub 1$ PER IL CALCOLO DEL FLUSSO TERMICO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daneri, A.; Daneri, A.

1964-01-01

The program DESTHEC DP, in FORTRAN MONITOR for the IBM 7090, solves the transport equation for thermal neutrons in slab geometry. For the energy, Galerkin's method with the double P/sub 1/ approximation is used, Comparison shows good agreement between DESTHEC DP results and results obtained by the THERMOS program, which solves the transport equation in integral form. The theory is presented, and input and output are discussed. Numerical results are included, as well as the program listing. (D.C.W.)

78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... correspondence to addressees and subscribers through a computer-based email distribution system. Since then, the... Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this... available operating reactor licensing correspondence, effective December 9, 2013. Official agency records...

76 FR 14436 - University of Wisconsin, University of Wisconsin Nuclear Reactor; Notice of Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

..., University of Wisconsin Nuclear Reactor; Notice of Issuance of Environmental Assessment and Finding of No... operation of the University of Wisconsin Nuclear Reactor. This action is necessary to add supplemental... of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

Special Analysis for the Disposal of the INL Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) Waste Stream at the Area 5 Radioactive Waste Management Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shott, Gregory

This special analysis (SA) evaluates whether the Idaho National Laboratory (INL) Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) waste stream (INEL167203QR1, Revision 0) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Waste Associated with the Unirradiated LWBR waste meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P performance objectives (DOE 1999). The INL Waste Associated with the Unirradiated LWBR waste stream is recommended for acceptance with the conditionmore » that the total uranium-233 ( 233U) inventory be limited to 2.7E13 Bq (7.2E2 Ci).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grabaskas, David; Bucknor, Matthew; Jerden, James

A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gapsmore » in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.« less

Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

NASA Astrophysics Data System (ADS)

Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

2015-12-01

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

Status and progress of the RERTR program in the year 2000.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

2000-09-28

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellentmore » irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm{sup 3} range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm{sup 3} are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm{sup 3} is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission {sup 99}Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO{sub 2} dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-11reactor in Berlin, Germany, was completed and conversion of the La Reins reactor in Santiago, Chile, began. These are exciting times for the program. In the fuel development area, the RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling fi.uther conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The {sup 99}Mo effort has reached the point where it appears feasible for all the {sup 99}Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

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 higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could actually help extend the life of a fuel pin in a "long pin" reactor design to a higher peak fuel burnup.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

Validation of IRDFF in 252Cf standard and IRDF-2002 reference neutron fields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simakov, Stanislav; Capote Noy, Roberto; Greenwood, Lawrence R.

The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f.) and reference 235U(nth,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm the recommended spectrum for 252Cf spontaneous fission source; that was not the case for the current recommended spectra for 235U(nth,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF- 2002 collection. Before this analysis, the ISFN spectrum was resimulated to remove unphysical oscillations in spectrum. IRDFF-1.03 was shown to reasonably reproducemore » the spectrum-averaged data measured in these fields except for the case of YAYOI.« less

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.

1991-12-31

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

75 FR 71744 - Toshiba Corporation Power Systems Company Notice of Receipt and Availability of an Application...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-24

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0361] Toshiba Corporation Power Systems Company Notice of Receipt and Availability of an Application for Renewal of the U.S. Advanced Boiling Water Reactor Design... application for a design certification (DC) renewal for the U.S. Advanced Boiling Water Reactor (ABWR). An...

RERTR-12 Insertion 2 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez; G. S. Chang; D. M. Wachs

2012-09-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

SoLid Detector Technology

NASA Astrophysics Data System (ADS)

Labare, Mathieu

2017-09-01

SoLid is a reactor anti-neutrino experiment where a novel detector is deployed at a minimum distance of 5.5 m from a nuclear reactor core. The purpose of the experiment is three-fold: to search for neutrino oscillations at a very short baseline; to measure the pure 235U neutrino energy spectrum; and to demonstrate the feasibility of neutrino detectors for reactor monitoring. This report presents the unique features of the SoLid detector technology. The technology has been optimised for a high background environment resulting from low overburden and the vicinity of a nuclear reactor. The versatility of the detector technology is demonstrated with a 288 kg detector prototype which was deployed at the BR2 nuclear reactor in 2015. The data presented includes both reactor on, reactor off and calibration measurements. The measurement results are compared with Monte Carlo simulations. The 1.6t SoLid detector is currently under construction, with an optimised design and upgraded material technology to enhance the detector capabilities. Its deployement on site is planned for the begin of 2017 and offers the prospect to resolve the reactor anomaly within about two years.

Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

NASA Astrophysics Data System (ADS)

Towell, Rusty; Niffte Collaboration

2015-10-01

The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

75 FR 66168 - Seeks Qualified Candidates for the Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-27

... NUCLEAR REGULATORY COMMISSION Seeks Qualified Candidates for the Advisory Committee on Reactor... Reactor Safeguards (ACRS). Submit r[eacute]sum[eacute]s to Ms. Brandi Hamilton, ACRS, Mail Stop T2E-26, U... of existing and proposed nuclear power plants and on the adequacy of proposed reactor safety...

Measurement of the 23Na(n,2n) cross section in 235U and 252Cf fission neutron spectra

NASA Astrophysics Data System (ADS)

Košťál, Michal; Schulc, Martin; Rypar, Vojtěch; Losa, Evžen; Švadlenková, Marie; Baroň, Petr; Jánský, Bohumil; Novák, Evžen; Mareček, Martin; Uhlíř, Jan

2017-09-01

The presented paper aims to compare the calculated and experimental reaction rates of 23Na(n,2n)22Na in a well-defined reactor spectra and in the spontaneous fission spectrum of 252Cf. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination.Estimation of this cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of long-term activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. In the case of reactor spectrum, reasonable agreement was not achieved with any library. However, in the case of 252Cf spectrum agreement was achieved with IRDFF, JEFF-3.1 and JENDL libraries.

Elastoplasticidad anisotropa de metales en grandes deformaciones

NASA Astrophysics Data System (ADS)

Caminero Torija, Miguel Angel

El objetivo de este trabajo es el desarrollo de modelos y algoritmos numericos que simulen el comportamiento del material bajo estas condiciones en el contexto de programas de elementos finitos, dando como resultado predicciones mas precisas de los procesos de conformado y deformacion plastica en general. Para lograr este objetivo se han desarrollado diversas tareas destinadas a mejorar las predicciones en tres aspectos fundamentales. El primer aspecto consiste en la mejora de la descripcion del endurecimiento cinematico anisotropo en pequenas deformaciones, lo cual se ha realizado a traves de modelos y algoritmos implicitos de superficies multiples. Ha sido estudiada la consistencia de este tipo de modelos tanto si estan basados en una regla implicita similar a la de Mroz o en la regla de Prager. Ademas se han simulado los ensayos de Lamba y Sidebottom, obteniendo, en contra de la creencia general, muy buenas predicciones con la regla de Prager. Dichos modelos podrian ser extendidos de forma relativamente facil para considerar grandes deformaciones a traves de procedimientos en deformaciones logaritmicas, similares a los desarrollados en esta tesis y detallados a continuacion. El segundo aspecto consiste en la descripcion de la anisotropia elastoplastica inicial. Esto se ha conseguido mediante el desarrollo de modelos y algoritmos para plasticidad anisotropa en grandes deformaciones, bien ignorando la posible anisotropia elastica, bien considerandola simultaneamente con la anisotropia plastica. Para ello ha sido necesario desarrollar primero un nuevo algoritmo de elastoplasticidad anisotropa en pequenas deformaciones consistentemente linealizado y sin despreciar ningun termino, de tal forma que se conserve la convergencia cuadratica de los metodos de Newton. Este algoritmo en pequenas deformaciones ha servido para realizar la correccion plastica de dos algoritmos en grandes deformaciones. El primero de estos algoritmos es una variacion del clasico algoritmo de Eterovic y Bathe para incluir la posibilidad de plasticidad anisotropa con endurecimiento mixto. Este primer algoritmo esta restringido a casos de isotropia elastica. La isotropia elastica es una hipotesis bastante habitual en plasticidad anisotropa y tiene la ventaja de que permite el uso de formulaciones mixtas u/p. El segundo algoritmo, mas complejo y general, incluye la posibilidad de elasticidad anisotropa, plasticidad anisotropa y endurecimiento mixto. Este algoritmo supone una contribucion importante ya que esta basado en hipotesis comunmente aceptadas y utilizadas en elastoplasticidad isotropa: descomposicion multiplicativa del gradiente de deformaciones en parte elastica y parte plastica, descripcion hiperelastica sencilla en funcion de deformaciones logaritmicas e integracion exponencial que conserva el volumen. Ademas, la estructura final del algoritmo es modular y relativamente sencilla, consistiendo en un pre- y un postprocesador geometrico y una correccion plastica realizada en pequenas deformaciones. El algoritmo esta consistentemente linealizado para conservar la convergencia cuadratica asintotica de los metodos de Newton y la forma final que toma dicha linealizacion es similar al caso de isotropia elastoplastica implementado; consiste en el modulo tangente algoritmico de pequenas deformaciones sobre el que se aplica una transformacion para convertirlo en el de grandes deformaciones. Todos estos modelos han sido implementados en un codigo propio de elementos finitos denominado DULCINEA, el cual tiene formulaciones lagrangianas totales y actualizadas para grandes deformaciones. Una de las tareas necesarias para poder realizar las simulaciones, ha sido el estudio e implementacion de diferentes elementos que no sufran el bloqueo volumetrico severo que se observa en formulaciones estandar basadas en desplazamientos. Este bloqueo se debe a la condicion de quasi-incompresibilidad que imponen los modelos de plasticidad desviadores y consiste en una respuesta exageradamente rigida de la solucion obtenida por el metodo de los elementos finitos estandar. Entre los elementos implementados cabe destacar el basado en la formulacion mixta u/p, que contiene una interpolacion adicional de grados de libertad de presion. Estos grados de libertad adicionales habitualmente son internos al elemento en mecanica de solidos. En este trabajo se ha desarrollado e implementado en DULCINEA una familia de elementos tridimensionales mixtos en grandes deformaciones que incluye el caso particular BMIX 27/27/4, basado en la formulacion u/p, constituido por 27 nudos, con 27 puntos de integracion estandar y 4 grados de libertad de presiones, y que pasa la condicion Inf-Sup o de Babuska-Brezzi. Sin embargo, se ha observado que la formulacion u/p presenta ciertas limitaciones bajo las hipotesis conjuntas de anisotropia elastica y anisotropia plastica. (Abstract shortened by UMI.)

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated inmore » the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Jaluvka, D.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclearmore » Security Administration’s Office of Material Management and Minimization (M 3).« less

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hursin, M.; Perret, G.

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handlingmore » and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)« less

Continuous production of pectinase by immobilized yeast cells on spent grains.

PubMed

Almeida, Catarina; Brányik, Tomás; Moradas-Ferreira, Pedro; Teixeira, José

2003-01-01

A yeast strain secreting endopolygalacturonase was used in this work to study the possibility of continuous production of this enzyme. It is a feasible and interesting alternative to fungal batch production essentially due to the specificity of the type of pectinase excreted by Kluyveromyces marxianus CCT 3172, to the lower broth viscosity and to the easier downstream operations. In order to increase the reactors' productivity, a cellulosic carrier obtained from barley spent grains was tested as an immobilization support. Two types of reactors were studied for pectinase production using glucose as a carbon and energy source--a continuous stirred tank reactor (CSTR) and a packed bed reactor (PBR) with recycled flow. The highest value for pectinase volumetric productivity (P(V)=0.98 U ml(-1) h(-1)) was achieved in the PBR for D=0.40 h(-1), a glucose concentration on the inlet of S(in)=20 g l(-1), and a biomass load in the support of X(i)=0.225 g g(-1). The results demonstrate the attractiveness of the packed bed system for pectinase production.

Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

2016-06-01

Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

Method for improving performance of irradiated structural materials

DOEpatents

Megusar, Janez; Harling, Otto K.; Grant, Nicholas J.

1989-01-01

Method for extending service life of nuclear reactor components prepared from ductile, high strength crystalline alloys obtained by devitrification of metallic glasses. Two variations of the method are described: (1) cycling the temperature of the nuclear reactor between the operating temperature which leads to irradiation damage and a l The U.S. Government has rights in this invention by virtue of Department of Energy, Office of Fusion Energy, Grant No. DE-AC02-78ER-10107.

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

A roadmap for nuclear energy technology

NASA Astrophysics Data System (ADS)

Sofu, Tanju

2018-01-01

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA's energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending "retirement cliff" of the existing U.S. nuclear fleet, representing over 60 percent of the nation's emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and "once-through" fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.4 Communications. Except where otherwise... Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission...

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.4 Communications. Except where otherwise... Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission...

Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

NASA Technical Reports Server (NTRS)

Clement, J. D.

1973-01-01

Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission... 10 Energy 2 2012-01-01 2012-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.4 Communications. Except where otherwise...

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

The role of mass spectrometry to study the Oklo-Bangombé natural reactors.

PubMed

De Laeter, J R; Hidaka, H

2007-01-01

The discovery of the existence of chain reactions at the Oklo natural reactors in Gabon, Central Africa in 1972 was a triumph for the accuracy of mass spectrometric measurements, in that a 0.5% anomaly in the (235)U/(238)U ratio of certain U ore samples indicated a depletion in (235)U. Mass spectrometric techniques thereafter played a dominant role in determining the nuclear parameters of the reactor zones themselves, and in deciphering the geochemical characteristics of various elements in the U-rich ore and in the surrounding rock strata. The variations in the isotopic composition of a large number of elements, caused by a combination of nuclear fission, neutron capture and radioactive decay, provide a powerful tool for investigating this unique geological environment. Mass spectrometry can be used to measure the present-day elemental and isotopic abundances of numerous elements, so as to decipher the past history of the reactors and examine the retentivity/mobility of these elements. Many of the fission products have a radioactive decay history that have been used to date the age and duration of the reactor zones, and to provide insight into their nuclear and geochemical behavior as a function of time. The Oklo fission reactors and their near neighbor at Bangombé, some 30 km to the south-east of Oklo, are unique in that not only did they become critical some 2 x 10(9) years ago, but also the deposits have been preserved over this period of geological time. The long-term geochemical behavior of actinides and fission products have been extensively studied by a variety of mass spectrometric techniques over the past 30 years to provide us with significant information on the mobility/retentivity of this material in a natural geological repository. The Oklo-Bangombé natural reactors are therefore geological analogs that can be evaluated in terms of possible radioactive waste containment sites. As more reactor zones were discovered, it was realized that they could be classified into two groups according to their burial depth in the Oklo mine-site. Reactor Zones 10, 13, and 16 were buried more deeply, and were therefore less weathered than the other zones. The less-weathered zones are of great importance in mobility/retentivity studies and therefore to the question of radioactive waste containment. Isotopic studies of these natural reactors are also of value in physics to examine possible variations in fundamental constants over the past 2 billion years.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gehin, Jess C.; Powers, Jeffrey J.

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing thorium-based MSRs.« less

Future Scenarios for Fission Based Reactors

NASA Astrophysics Data System (ADS)

David, S.

2005-04-01

The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile inventory, capacity to be deployed, induced radiotoxicities, and R&D efforts.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Bernard, D.; Dos Santos, N.

This paper describes the method to define relevant targeted integral measurements that allow the improvement of nuclear data evaluations and the determination of corresponding reliable covariances. {sup 235}U and {sup 56}Fe examples are pointed out for the improvement of JEFF3 data. Utilizations of these covariances are shown for Sensitivity and Representativity studies, Uncertainty calculations, and Transposition of experimental results to industrial applications. S/U studies are more and more used in Reactor Physics and Safety-Criticality. However, the reliability of study results relies strongly on the ND covariance relevancy. Our method derives the real uncertainty associated with each evaluation from calibration onmore » targeted integral measurements. These realistic covariance matrices allow reliable JEFF3.1.1 calculation of prior uncertainty due to nuclear data, as well as uncertainty reduction based on representative integral experiments, in challenging design calculations such as GEN3 and RJH reactors.« less

PROSPECT - A Precision Oscillation and Spectrum Experiment

NASA Astrophysics Data System (ADS)

Zhang, Xianyi; Prospect Collaboration

2017-01-01

PROSPECT, the PRecision Oscillation and SPECTrum Experiment, is a multi-phased short baseline reactor antineutrino experiment that aims to precisely measure the U-235 antineutrino spectrum and prob for oscillation effects involving a possible Δm2 1 eV2 scale sterile neutrino. In PROSPECT Phase-I, an optically segmented Li-6 loaded liquid scintillator detector will be deployed at at the baseline of 7-12m from the High Flux Isotope Reactor at the Oak Ridge National Laboratory. PROSPECT will measure the spectrum of U-235 to aid in resolving the unexplained inconsistency between predictive spectral models and recent experimental measurements using LEU cores, while the oscillation measurement will probe the best fit region suggested by global fitting studies within 1-year data taking. This talk will introduce the design of PROSPECT Phase-I, the discovery potential of the experiment, and the progress the collaboration has made toward realizing PROSPECT Phase-I. Department of Energy

Kinetic modelling and microbial community assessment of anaerobic biphasic fixed film bioreactor treating distillery spent wash.

PubMed

Acharya, Bhavik K; Pathak, Hilor; Mohana, Sarayu; Shouche, Yogesh; Singh, Vasdev; Madamwar, Datta

2011-08-01

Anaerobic digestion, microbial community structure and kinetics were studied in a biphasic continuously fed, upflow anaerobic fixed film reactor treating high strength distillery wastewater. Treatment efficiency of the bioreactor was investigated at different hydraulic retention times (HRT) and organic loading rates (OLR 5-20 kg COD m⁻³ d⁻¹). Applying the modified Stover-Kincannon model to the reactor, the maximum removal rate constant (U(max)) and saturation value constant (K(B)) were found to be 2 kg m⁻³ d⁻¹ and 1.69 kg m⁻³ d⁻¹ respectively. Bacterial community structures of acidogenic and methanogenic reactors were assessed using culture-independent analyses. Sequencing of 16S rRNA genes exhibited a total of 123 distinct operational taxonomic units (OTUs) comprising 49 from acidogenic reactor and 74 (28 of eubacteria and 46 of archaea) from methanogenic reactor. The findings reveal the role of Lactobacillus sp. (Firmicutes) as dominant acid producing organisms in acidogenic reactor and Methanoculleus sp. (Euryarchaeotes) as foremost methanogens in methanogenic reactor. Copyright © 2011 Elsevier Ltd. All rights reserved.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

NASA Astrophysics Data System (ADS)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

White Paper – Use of LEU for a Space Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin; Mcclure, Patrick Ray

Historically space reactors flown or designed for the U.S. and Russia used Highly Enriched Uranium (HEU) for fuel. HEU almost always produces a small and lighter reactor. Since mass increases launch costs or decreases science payloads, HEU was the natural choice. However in today’s environment, the proliferation of HEU has become a major concern for the U.S. government and hence a policy issue. In addition, launch costs are being reduced as the space community moves toward commercial launch vehicles. HEU also carries a heavy security cost to process, test, transport and launch. Together these issues have called for a re-investigationmore » into space reactors the use Low Enriched Uranium (LEU) fuel.« less

Biodegradation of Jet Fuel-4 (JP-4) in Sequencing Batch Reactors

DTIC Science & Technology

1992-06-01

nalw~eo %CUMENTATION PAGE__ _ _ _ _ _ _ _ _O 74S Ab -A258 020 L AW POi~W6 DATI .~ TYP AIMqm ,-& 0 U. glbs A~ I ma"&LFUN Mu BIODEGRADATION OF JET FUEL...Specific Objectives of This Proposal Are: 1. To assess the ability of C. resinae , P. chrysosporium and selected bacterial consortia to degrade individual...chemical components of JP-4. 2. To develop a sequencing batch reactor that utilizes C. resinae to degrade chemical components of JP-4 in contaminated

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Ye, Bei; Mei, Zhigang

Uranium silicide (U 3Si 2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U 3Si 2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U 3Si 2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuelmore » material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U 3Si 2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U 3Si 2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U 3Si 2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U 3Si 2 fuel as an accident-tolerant alternative for uranium dioxide.« less

U.S./CIS eye joint nuclear rocket venture

NASA Technical Reports Server (NTRS)

Clark, John S.; Mcilwain, Melvin C.; Smetanikov, Vladimir; D'Yakov, Evgenij K.; Pavshuk, Vladimir A.

1993-01-01

An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

NASA Astrophysics Data System (ADS)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

ERIC Educational Resources Information Center

National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

METHOD OF OPERATING NUCLEAR REACTORS

DOEpatents

Untermyer, S.

1958-10-14

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Michael A. Pope; Harold F. McFarlane

2012-11-01

The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactormore » (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.« less

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactormore » physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.« less

Nuclear instrumentation in VENUS-F

NASA Astrophysics Data System (ADS)

Wagemans, J.; Borms, L.; Kochetkov, A.; Krása, A.; Van Grieken, C.; Vittiglio, G.

2018-01-01

VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).

Proceedings of the Discussion Meeting on Thermodynamics of Alloys Held in Barcelona, Spain on 23-26 May 1999. Series B. Volume 86

DTIC Science & Technology

1990-05-01

PUBLICACION DE LA REAL SOCIEOAD ESPANOLA DE PISICA AD-A 24 1 031 ANALES DE FISICA SERI[ B -Dhf 4!S - -c) D- NA- DZ-7 3. APLICACIONES , METODOS. -- E...ANALES DE FISICA Serie A: Fen6menos e Interacciones Serie B: Aplicaciones , M~todos e Instrumentos PUBLICACION DE LA REAL SOCIEDAD ESPANOLA DE FISICA...Apartado 1065. Avenida Reina Mercedes, s/n. Universidad de Sevilla. Sevilla - 41080. EDITOR DE FISICA ATOMICA, MOLECULAR Y NUCLEAR Prof. D. Jos6 Maria

Methanosarcina Play an Important Role in Anaerobic Co-Digestion of the Seaweed Ulva lactuca: Taxonomy and Predicted Metabolism of Functional Microbial Communities.

PubMed

FitzGerald, Jamie A; Allen, Eoin; Wall, David M; Jackson, Stephen A; Murphy, Jerry D; Dobson, Alan D W

2015-01-01

Macro-algae represent an ideal resource of third generation biofuels, but their use necessitates a refinement of commonly used anaerobic digestion processes. In a previous study, contrasting mixes of dairy slurry and the macro-alga Ulva lactuca were anaerobically digested in mesophilic continuously stirred tank reactors for 40 weeks. Higher proportions of U. lactuca in the feedstock led to inhibited digestion and rapid accumulation of volatile fatty acids, requiring a reduced organic loading rate. In this study, 16S pyrosequencing was employed to characterise the microbial communities of both the weakest (R1) and strongest (R6) performing reactors from the previous work as they developed over a 39 and 27-week period respectively. Comparing the reactor communities revealed clear differences in taxonomy, predicted metabolic orientation and mechanisms of inhibition, while constrained canonical analysis (CCA) showed ammonia and biogas yield to be the strongest factors differentiating the two reactor communities. Significant biomarker taxa and predicted metabolic activities were identified for viable and failing anaerobic digestion of U. lactuca. Acetoclastic methanogens were inhibited early in R1 operation, followed by a gradual decline of hydrogenotrophic methanogens. Near-total loss of methanogens led to an accumulation of acetic acid that reduced performance of R1, while a slow decline in biogas yield in R6 could be attributed to inhibition of acetogenic rather than methanogenic activity. The improved performance of R6 is likely to have been as a result of the large Methanosarcina population, which enabled rapid removal of acetic acid, providing favourable conditions for substrate degradation.

On similarity of various reactor spectra and 235U prompt fission neutron spectrum.

PubMed

Košťál, Michal; Matěj, Zdeněk; Losa, Evžen; Huml, Ondřej; Štefánik, Milan; Cvachovec, František; Schulc, Martin; Jánský, Bohumil; Novák, Evžen; Harutyunyan, Davit; Rypar, Vojtěch

2018-05-01

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235 U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 - 6 MeV. Above this energy the reactor spectra becomes identical with the 235 U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra. Copyright © 2018 Elsevier Ltd. All rights reserved.

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vins, M.

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe.more » Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)« less

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Totev, T.; Stevens, J.; Kim, Y. S.

2010-03-01

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 Testmore » 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.« less

76 FR 70778 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-15

... Pike, Rockville, Maryland. Thursday, December 1, 2011, Conference Room T2-B1, 11545 Rockville Pike... subsequent COL application for Levy County, Units 1 and 2, and the NRC staff's associated safety evaluation... pursuant to 5 U.S.C. 552b(c)(4).] Friday, December 2, 2011, Conference Room T2-B1, 11545 Rockville Pike...

Small space reactor power systems for unmanned solar system exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

75 FR 3501 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on February 4-6, 2010, 11545 Rockville Pike...

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

75 FR 12311 - FirstEnergy Nuclear Operating Company; Notice of Consideration of Issuance of Amendment to...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-15

... INFORMATION CONTACT: Michael Mahoney, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission... Licensing Branch III-2, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR... NUCLEAR REGULATORY COMMISSION [Docket No. 50-346] FirstEnergy Nuclear Operating Company; Notice of...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bi, G.; Liu, C.; Si, S.

This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis ofmore » reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared with reference full UOX core. The fuel cycle analysis has shown that {sup 233}U mono-recycling with U{sub 3}ThOX fuel could save 13% of natural uranium resource compared with UOX once through fuel cycle, slightly more than that of Plutonium single-recycling with MOX fuel. If {sup 233}U multi-recycling with U{sub 3}ThOX fuel is implemented, more natural uranium resource would be saved. (authors)« less

Progress of the RERTR program in 2001.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

2002-03-07

This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualificationmore » of the U-Mo dispersion fuels has been delayed by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid-2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid-2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIINM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4,562 spent fuel assemblies from foreign research reactors had been received by that date by the U.S. under the FRR SNF acceptance policy. The RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling further conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the U.S. FRR SNF Acceptance Program. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

NASA Astrophysics Data System (ADS)

Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

2016-03-01

The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant

PubMed Central

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C.; Utsunomiya, Satoshi

2017-01-01

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0–3.4 μm) comprise SiO2 glass matrices and ~10-nm-sized Zn–Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1–19 wt% Cs as Cs2O). Trace amounts of U are also associated with the Zn–Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP. PMID:28198440

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C.; Utsunomiya, Satoshi

2017-02-01

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0-3.4 μm) comprise SiO2 glass matrices and ~10-nm-sized Zn-Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1-19 wt% Cs as Cs2O). Trace amounts of U are also associated with the Zn-Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP.

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant.

PubMed

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C; Utsunomiya, Satoshi

2017-02-15

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0-3.4 μm) comprise SiO 2 glass matrices and ~10-nm-sized Zn-Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1-19 wt% Cs as Cs 2 O). Trace amounts of U are also associated with the Zn-Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO 2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP.

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

Metodo de Archivar las Observaciones del Comportamiento del Nino, Como Guia para Entenderlo Mejor (Methods of Recording Observations of Children's Behavior, A Guide for Better Understanding)

ERIC Educational Resources Information Center

Stamp, Isla M.

1971-01-01

Copies of the Behaviour Study Technique described in this article may be obtained in English from the Australian Council for Educational Research, Frederick St., Hawthorn, Victoria, Australia 3122. (RY)

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variablemore » lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strickland, Christopher E.; Lawter, Amanda R.; Qafoku, Nikolla

Isotopes of iodine were generated during plutonium production from nine production reactors at the U.S. Department of Energy Hanford Site. The long half-life 129I generated at the Hanford Site during reactor operations was 1) stored in single-shell and double-shell tanks, 2) discharged to liquid disposal sites (e.g., cribs and trenches), 3) released to the atmosphere during fuel reprocessing operations, or 4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129I to the subsurface have resulted in several large, though dilute, plumes in the groundwater, including the plume in the 200-UP-1more » operable unit. There is also 129I remaining in the vadose zone beneath disposal or leak locations. Because 129I is an uncommon contaminant, relevant remediation experience and scientific literature are limited.« less

Impact of fission neutron energies on reactor antineutrino spectra

NASA Astrophysics Data System (ADS)

Littlejohn, B. R.; Conant, A.; Dwyer, D. A.; Erickson, A.; Gustafson, I.; Hermanek, K.

2018-04-01

Recent measurements of reactor-produced antineutrino fluxes and energy spectra are inconsistent with models based on measured thermal fission beta spectra. In this paper, we examine the dependence of antineutrino production on fission neutron energy. In particular, the variation of fission product yields with neutron energy has been considered as a possible source of the discrepancies between antineutrino observations and models. In simulations of low-enriched and highly-enriched reactor core designs, we find a substantial fraction of fissions (from 5% to more than 40%) are caused by nonthermal neutrons. Using tabulated evaluations of nuclear fission and decay, we estimate the variation in antineutrino emission by the prominent fission parents U 235 , Pu 239 , and Pu 241 versus neutron energy. The differences in fission neutron energy are found to produce less than 1% variation in detected antineutrino rate per fission of U 235 , Pu 239 , and Pu 241 . Corresponding variations in the antineutrino spectrum are found to be less than 10% below 7 MeV antineutrino energy, smaller than current model uncertainties. We conclude that insufficient modeling of fission neutron energy is unlikely to be the cause of the various reactor anomalies. Our results also suggest that comparisons of antineutrino measurements at low-enriched and highly-enriched reactors can safely neglect the differences in the distributions of their fission neutron energies.

Convective cooling in a pool-type research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

2016-01-22

A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.

2015-12-15

A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields inmore » the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.« less

Kinetic of carbonaceous substrate in an upflow anaerobic sludge sludge blanket (UASB) reactor treating 2,4 dichlorophenol (2,4 DCP).

PubMed

Sponza, Delia Teresa; Uluköy, Ayşen

2008-01-01

The performance of an upflow anaerobic sludge blanket (UASB) reactor treating 2,4 dichlorophenol (2,4 DCP) was evaluated at different hydraulic retention times (HRTs) using synthetic wastewater in order to obtain the growth substrate (glucose-COD) and 2,4 DCP removal kinetics. Treatment efficiencies of the UASB reactor were investigated at different hydraulic retention times (2-20 h) corresponding to a food to mass (F/M) ratio of 1.2-1.92 g-COD g(-1) VSS day(-1). A total of 65-83% COD removal efficiencies were obtained at HRTs of 2-20 h. In all, 83% and 99% 2,4 DCP removals were achieved at the same HRTs in the UASB reactor. Conventional Monod, Grau Second-order and Modified Stover-Kincannon models were applied to determine the substrate removal kinetics of the UASB reactor. The experimental data obtained from the kinetic models showed that the Monod kinetic model is more appropriate for correlating the substrate removals compared to the other models for the UASB reactor. The maximum specific substrate utilization rate (k) (mg-COD mg(-1) SS day(-1)), half-velocity concentration (K(s)) (mg COD l(-1)), growth yield coefficient (Y) (mg mg(-1)) and bacterial decay coefficient (b) (day(-1)) were 0.954 mg-COD mg(-1) SS day(-1), 560.29 mg-COD l(-1), 0.78 mg-SS g(-1)-COD, 0.093 day(-1) in the Conventional Monod kinetic model. The second-order kinetic coefficient (k(2)) was calculated as 0.26 day(-1) in the Grau reaction kinetic model. The maximum COD removal rate constant (U(max)) and saturation value (K(B)) were calculated as 7.502 mg CODl(-1)day(-1) and 34.56 mg l(-1)day(-1) in the Modified Stover-Kincannon Model. The (k)(mg-2,4 DCP mg(-1) SS day(-1)), (K(s)) (mg 2,4 DCPl(-1)), (Y) (mg SS mg(-1) 2,4 DCP) and (k(d)) (day(-1)) were 0.0041 mg-2,4 DCP mg(-1) SS day(-1), 2.06 mg-COD l(-1), 0.0017 mg-SS mg(-1) 2,4 DCP and 3.1 x 10(-5) day(-1) in the Conventional Monod kinetic model for 2,4 DCP degradation. The second-order kinetic coefficient (k(2)) was calculated as 0.30 day(-1) in the Grau reaction kinetic model. The maximum 2,4 DCP removal rate constant (U(max)) and saturation value (K(B)) were calculated as 0.01 mg COD l(-1) day(-1) and 9.8 x 10(-3) mg l(-1) day(-1) in the Modified Stover-Kincannon model.

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, James J.; Grandy, Christopher

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. Themore » various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.« less

Beta decay heat following U-235, U-238 and Pu-239 neutron fission

NASA Astrophysics Data System (ADS)

Li, Shengjie

1997-09-01

This is an experimental study of beta-particle decay heat from 235U, 239Pu and 238U aggregate fission products over delay times 0.4-40,000 seconds. The experimental results below 2s for 235U and 239Pu, and below 20s for 238U, are the first such results reported. The experiments were conducted at the UMASS Lowell 5.5-MV Van de Graaff accelerator and 1-MW swimming-pool research reactor. Thermalized neutrons from the 7Li(p,n)7Be reaction induced fission in 238U and 239Pu, and fast neutrons produced in the reactor initiated fission in 238U. A helium-jet/tape-transport system rapidly transferred fission fragments from a fission chamber to a low background counting area. Delay times after fission were selected by varying the tape speed or the position of the spray point relative to the beta spectrometer that employed a thin-scintillator-disk gating technique to separate beta-particles from accompanying gamma-rays. Beta and gamma sources were both used in energy calibration. Based on low-energy(<1 MeV) internal-conversion electron studies, a set of trial responses for the spectrometer was established and spanned electron energies 0-10 MeV. Measured beta spectra were unfolded for their energy distributions by the program FERD, and then compared to other measurements and summation calculations based on ENDF/B-VI fission-product data performed on the LANL Cray computer. Measurements of the beta activity as a function of decay time furnished a relative normalization. Results for the beta decay heat are presented and compared with other experimental data and the summation calculations.

Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

NASA Astrophysics Data System (ADS)

Bhattacharya, Sumit

High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the metal organic chemical precursors tetrakis dimethylamido zirconium (TDMAZr) and ammonia( NH3) for succesful deposition of ZrN coating. Nitrogen (N2) gas carried the chemicals to a hot wall reactor maintained at a temperature range of 235 to 245 °C. The ALD system design evolved over the course of this research as the process variables were steadily improved. The conditions found deemed for attaining best coating were at a temperature of 245 °C, with pulse time of 0.8 seconds for TDMAZr and 0.1 seconds for NH3 along with 15 seconds of purge time in-between each cycle. The ALD system was successful in making 1-micrometer (um) ZrN with low levels of chemical impurities over U-Mo powder batches. The deposited coatings were characterized using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), electron energy loss spectroscopy (EELS) and Transmission electron microscope (TEM). This document describes the establishment of the Savannah 200 ALD system, precursor surface reaction procedures and finally the nature of the coating achieved, including characterization of the coating at the different stages of deposition. It was found that an interlayer of alumina in between ZrN and the U-Mo surface was required to reduce the residual stress generated during the ALD procedure. The alumina not only removed the risk of cracking and spallation of the ZrN coating but also provided adequate strength for the barrier layer to withstand the fuel plate rolling conditions. The ZrN coating was nano crystalline in nature, with grain size varying from 5-10 nm, the deposited layer was found to be dense consisting of a layered structure. The coating could retain its crystallinity and maintain its phase when irradiated with 1 MeV single charged ion Kr to produce a damage of 10 displacement per atom (DPA) at intermediate voltage electron microscopy (IVEM).

Risk Management for Sodium Fast Reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event withmore » the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.« less

Packed bed reactor for photochemical .sup.196 Hg isotope separation

DOEpatents

Grossman, Mark W.; Speer, Richard

1992-01-01

Straight tubes and randomly oriented pieces of tubing having been employed in a photochemical mercury enrichment reactor and have been found to improve the enrichment factor (E) and utilization (U) compared to a non-packed reactor. One preferred embodiment of this system uses a moving bed (via gravity) for random packing.

DYNAMIC AND STATIC PARAMETERS OF THE AQUEOUS HOMOGENEOUS ARMOUR RESEARCH REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrell, C.W.; McElroy, W.N.

1959-06-01

A brief description of the aqueous homogeneous Armour Research Reactor is given. The negative reactivity coefficient resulting from a temperature increase was determined over a fuel temperature range of 37 to 150 deg F. Possession of an accurately calibrated rod and temperature coefficient permitted a direct measurement of the void coefficient. The reactor was taken to different power levels, and from the calibrated rod the total reduction in excess reactivity was obtained. During the power increase program additional U/sup 235/ and water were added to the core to determine the worth of U/sup 235/ and water. (W.D.M.)

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

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

1984-08-30

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

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, Paul R.; McLennan, George A.

1985-01-01

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

Il metodo di Aristarco à solo un modello?

NASA Astrophysics Data System (ADS)

Sigismondi, Costantino; Agolini, Giorgia

2017-06-01

The method of Aristarchus of Samo to measure the distance Earth-Sun is well known as the first step in the cosmological distance ladder. Our task is to understand if it is worth to repeat it also experimentally, with positive conclusions, especially with respect to the experimental design.

The Fukushima Nuclear Disaster and the U.S. Customs and Border Protection Response

NASA Astrophysics Data System (ADS)

McCormick, Kathy

2013-10-01

On 3/11/11, the reactors at the Fukushima Nuclear Plant in Japan were damaged by a magnitude 9.0 earthquake. Of the six reactors at the site, three were in operation prior to the event, and were automatically shut-down during the earthquake. Emergency cooling systems came online and were subsequently destroyed by a tsunami generated by the earthquake. For the operating reactors, all the reactor cores were exposed, resulting in overheating and the release of steam and hydrogen gas to the containment vessels, several of which subsequently exploded, releasing radioactivity into the atmosphere. The cores of the operating reactors melted down, and radioactive water was released to the ocean in cooling efforts. The primary radiation concerns in the United States from the disaster were radioactive plumes driven by westerly winds and contaminated commercial products and travelers. In the United States, one of the primary governmental organizations to respond to the disaster was U.S. Customs and Border Protection (CBP), which has responsibility to oversee the safety and security of cargo and travelers entering the United States. This talk will describe the various types of radioactive commodities and events encountered by CBP in the U.S. from the Fukushima disaster. Thanks to the CBP Teleforensics Center for their assistance with this presentation.

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moe, Wayne Leland

2016-08-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologiesmore » (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants« less

Bioprospecting of gum kondagogu (Cochlospermum gossypium) for bioremediation of uranium (VI) from aqueous solution and synthetic nuclear power reactor effluents.

PubMed

Sashidhar, R B; Selvi, S Kalaignana; Vinod, V T P; Kosuri, Tanuja; Raju, D; Karuna, R

2015-10-01

An ecofriendly green chemistry method using a natural biopolymer, Gum Kondagogu (GK) for the removal of U (VI) from aqueous, simulated nuclear effluents was studied. The adsorption characteristic of GK towards U (VI) from aqueous solution was studied at varied pH, contact time, adsorbent dose, initial U (VI) concentration and temperature using UV-Visible spectroscopy and ICP-MS. Maximum adsorption was seen at pH 4, 0.1% GK with 60 min contact time at room temperature. The GK- U (VI) composite was characterized by FT-IR, zeta potential, TEM and SEM-EDAX. The Langmuir isotherm was found to be 487 mg of U (VI) g(-1) of GK. The adsorption capacity and (%) of U (VI) was found to be 490 ± 5.4 mg g(-1) and 98.5%. Moreover adsorption of U (VI) by GK was not influenced by other cations present in the simulated effluents. The adsorbed U (VI) was efficiently stripped from composite using 1 M HCl. Copyright © 2015 Elsevier Ltd. All rights reserved.

Status and progress of the RERTR program in the year 2003.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

Phenomena Important in Molten Salt Reactor Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David J.; Brown, Nicholas R.; Denning, Richard

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

NASA Astrophysics Data System (ADS)

Terranova, Nicholas; Serot, Olivier; Archier, Pascal; De Saint Jean, Cyrille; Sumini, Marco

2017-09-01

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics

NASA Astrophysics Data System (ADS)

Moore, A. P.; Beeler, B.; Deo, C.; Baskes, M. I.; Okuniewski, M. A.

2015-12-01

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium-zirconium alloy (γ-U-Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U-Zr alloys. Uranium-rich U-Zr alloys (e.g. U-10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U-Pu-Zr and U-Zr-Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U-Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U-Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U-Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

NASA Astrophysics Data System (ADS)

Perez, Emmanuel; Sohn, Yong-Ho; Keiser, Dennis D.

2013-01-01

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

NASA Astrophysics Data System (ADS)

Azizov, E. A.; Gladush, G. G.; Dokuka, V. N.; Khayrutdinov, R. R.

2015-12-01

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

UF6 breeder reactor power plants for electric power generation

NASA Technical Reports Server (NTRS)

Rust, J. H.; Clement, J. D.; Hohl, F.

1976-01-01

The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

2017-01-12

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

Characterization of microstructural, mechanical and thermophysical properties of Th-52U alloy

NASA Astrophysics Data System (ADS)

Das, Santanu; Kaity, S.; Kumar, R.; Banerjee, J.; Roy, S. B.; Chaudhari, G. P.; Daniel, B. S. S.

2016-11-01

Th-52 wt.% U alloy has a microstructure featuring interspersed networks of uranium rich and thorium rich phases. Room temperature hardness of the alloy is more than twice that of unalloyed thorium. The alloy age hardens (550 °C) only slightly (peak hardness/hardness of solution heated and quenched = 1.05). Room temperature thermal conductivity (25.6 W m-1 °C-1) is close to that of uranium and most of the binary and ternary metallic alloy fuel materials. Average linear coefficient of thermal expansion (CTE) of Th-52 wt.% U alloy [11.2 × 10-06 °C-1 (27-290 °C) and 16.75 × 10-06 °C-1 (27-600 °C)] are comparable with that of many metallic alloy fuel candidates. Th-52 wt.% U alloy with non-age hardenable microstructure, appreciable thermal conductivity, moderate thermal expansion may find metallic fuel applications in nuclear reactors.

Reactors are indispensable for radioisotope production.

PubMed

Mushtaq, Ahmad

2010-12-01

Radioisotopes can be produced by reactors and accelerators. For certain isotopes there could be an advantage to a certain production method. However, nowadays many reports suggest, that useful isotopes needed in medicine, industry and research could be produced efficiently and dependence on reactors using enriched U-235 may be eliminated. In my view reactors and accelerators will continue to play their role side by side in the supply of suitable and economical sources of isotopes.

Determination of irradiated reactor uranium in soil samples in Belarus using 236U as irradiated uranium tracer.

PubMed

Mironov, Vladislav P; Matusevich, Janna L; Kudrjashov, Vladimir P; Boulyga, Sergei F; Becker, J Sabine

2002-12-01

This work presents experimental results on the distribution of irradiated reactor uranium from fallout after the accident at Chernobyl Nuclear Power Plant (NPP) in comparison to natural uranium distribution in different soil types. Oxidation processes and vertical migration of irradiated uranium in soils typical of the 30 km relocation area around Chernobyl NPP were studied using 236U as the tracer for irradiated reactor uranium and inductively coupled plasma mass spectrometry as the analytical method for uranium isotope ratio measurements. Measurements of natural uranium yielded significant variations of its concentration in upper soil layers from 2 x 10(-7) g g(-1) to 3.4 x 10(-6) g g(-1). Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 5 x 10(-12) g g(-1) to 2 x 10(-6) g g(-1) depending on the distance from Chernobyl NPP. In the majority of investigated soil profiles 78% to 97% of irradiated "Chernobyl" uranium is still contained in the upper 0-10 cm soil layers. The physical and chemical characteristics of the soil do not have any significant influence on processes of fuel particle destruction. Results obtained using carbonate leaching of 236U confirmed that more than 60% of irradiated "Chernobyl" uranium is still in a tetravalent form, ie. it is included in the fuel matrix (non-oxidized fuel UO2). The average value of the destruction rate of fuel particles determined for the Western radioactive trace (k = 0.030 +/- 0.005 yr(-1)) and for the Northern radioactive trace (k = 0.035 + 0.009 yr(-1)) coincide within experimental errors. Use of leaching of fission products in comparison to leaching of uranium for study of the destruction rate of fuel particles yielded poor coincidence due to the fact that use of fission products does not take into account differences in the chemical properties of fission products and fuel matrix (uranium).

Laser-induced breakdown spectroscopy of light water reactor simulated used nuclear fuel: Main oxide phase

DOE PAGES

Campbell, Keri R.; Judge, Elizabeth J.; Barefield, James E.; ...

2017-04-22

We show the analysis of light water reactor simulated used nuclear fuel using laser-induced breakdown spectroscopy (LIBS) is explored using a simplified version of the main oxide phase. The main oxide phase consists of the actinides, lanthanides, and zirconium. The purpose of this study is to develop a rapid, quantitative technique for measuring zirconium in a uranium dioxide matrix without the need to dissolve the material. A second set of materials including cerium oxide is also analyzed to determine precision and limit of detection (LOD) using LIBS in a complex matrix. Two types of samples are used in this study:more » binary and ternary oxide pellets. The ternary oxide, (U,Zr,Ce)O 2 pellets used in this study are a simplified version the main oxide phase of used nuclear fuel. The binary oxides, (U,Ce)O 2 and (U,Zr)O 2 are also examined to determine spectral emission lines for Ce and Zr, potential spectral interferences with uranium and baseline LOD values for Ce and Zr in a UO 2 matrix. In the spectral range of 200 to 800 nm, 33 cerium lines and 25 zirconium lines were identified and shown to have linear correlation values (R 2) > 0.97 for both the binary and ternary oxides. The cerium LOD in the (U,Ce)O 2 matrix ranged from 0.34 to 1.08 wt% and 0.94 to 1.22 wt% in (U,Ce,Zr)O 2 for 33 of Ce emission lines. The zirconium limit of detection in the (U,Zr)O 2 matrix ranged from 0.84 to 1.15 wt% and 0.99 to 1.10 wt% in (U,Ce,Zr)O 2 for 25 Zr lines. Finally, the effect of multiple elements in the plasma and the impact on the LOD is discussed.« less

Renewability and sustainability aspects of nuclear energy

NASA Astrophysics Data System (ADS)

Şahin, Sümer

2014-09-01

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, 233U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO2/RG-PuO2) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG-PuO2 + 96 % ThO2; 6 % RG-PuO2 + 94 % ThO2; 10 % RG-PuO2 + 90 % ThO2; 20 % RG-PuO2 + 80 % ThO2; 30 % RG-PuO2 + 70 % ThO2, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ˜ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ˜ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG-PuO2 fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MWth has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ˜160 kg 233U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ˜1.3.

ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

DOE PAGES

Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; ...

2016-05-01

Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

The Application of U-Np Fuel and {sup 6}Li Burnable Poison for Space Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nikitin, Konstantin L.; Saito, Masaki; Artisyuk, Vladimir V.

2003-11-15

The possible application of {sup 6}Li as a burnable poison and U-Np nitride as a fuel for space nuclear reactors has been studied. The analysis was performed for an infinite lattice with a leakage in the form of buckling and (U-Np)N fuel with 20% uranium enrichment. The combination of {sup 7}Li as a coolant and {sup 6}Li as a burnable poison results in a favorable criticality behavior during burnup. The parameters taken into consideration include the different fuel and coolant compositions, the form of absorber material, and the various absorber mass and concentrations. It was found that absorption properties ofmore » {sup 6}Li allow reaching the burnup value up to 67 GWd/tHM while reactivity swing is comparable with {beta}{sub eff}. The corresponding reactor lifetime is {approx}10 to 30 yr.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

78 FR 32279 - Advisory Committee On Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-29

... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards; Notice of Meeting In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on June 5-7, 2013, 11545 Rockville Pike...

76 FR 35922 - Interim Staff Guidance Regarding the Environmental Report for Applications To Construct and/or...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-20

... Nuclear Reactor Regulation on the information that should be included in the Environmental Report, which...: Mr. Scott Sloan, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-15227 Filed 6-17-11; 8:45 am] BILLING CODE 7590-01-P ...

Methanosarcina Play an Important Role in Anaerobic Co-Digestion of the Seaweed Ulva lactuca: Taxonomy and Predicted Metabolism of Functional Microbial Communities

PubMed Central

FitzGerald, Jamie A.; Allen, Eoin; Wall, David M.; Jackson, Stephen A.; Murphy, Jerry D.; Dobson, Alan D. W.

2015-01-01

Macro-algae represent an ideal resource of third generation biofuels, but their use necessitates a refinement of commonly used anaerobic digestion processes. In a previous study, contrasting mixes of dairy slurry and the macro-alga Ulva lactuca were anaerobically digested in mesophilic continuously stirred tank reactors for 40 weeks. Higher proportions of U. lactuca in the feedstock led to inhibited digestion and rapid accumulation of volatile fatty acids, requiring a reduced organic loading rate. In this study, 16S pyrosequencing was employed to characterise the microbial communities of both the weakest (R1) and strongest (R6) performing reactors from the previous work as they developed over a 39 and 27-week period respectively. Comparing the reactor communities revealed clear differences in taxonomy, predicted metabolic orientation and mechanisms of inhibition, while constrained canonical analysis (CCA) showed ammonia and biogas yield to be the strongest factors differentiating the two reactor communities. Significant biomarker taxa and predicted metabolic activities were identified for viable and failing anaerobic digestion of U. lactuca. Acetoclastic methanogens were inhibited early in R1 operation, followed by a gradual decline of hydrogenotrophic methanogens. Near-total loss of methanogens led to an accumulation of acetic acid that reduced performance of R1, while a slow decline in biogas yield in R6 could be attributed to inhibition of acetogenic rather than methanogenic activity. The improved performance of R6 is likely to have been as a result of the large Methanosarcina population, which enabled rapid removal of acetic acid, providing favourable conditions for substrate degradation. PMID:26555136

Performance of the MTR core with MOX fuel using the MCNP4C2 code.

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2016-08-01

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

76 FR 11289 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-01

... Thursday, October 21, 2010 (74 FR 65038-65039). Thursday, March 10, 2011, Conference Room T2-B1, 11545..., Units 1 and 2 extended power uprate application and the associated safety evaluation prepared by the NRC... Energy Point Beach, LLC, pursuant to 5 U.S.C. 552b(c)(4).] Friday, March 11, 2011, Conference Room T2-B1...

77 FR 30029 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-21

..., Rockville, Maryland. Wednesday, June 6, 2012, Conference Room T2-B1, 11545 Rockville Pike, Rockville... evaluation of the Fukushima Dai-ichi accident. 12:45 p.m.-2:15 p.m.: Proposed Revision 1 to Regulatory Guide... proprietary, pursuant to 5 U.S.C. 552b(c)(4)]. Thursday, June 7, 2012, Conference Room T2-B1, 11545 Rockville...

Hybrid systems for transuranic waste transmutation in nuclear power reactors: state of the art and future prospects

NASA Astrophysics Data System (ADS)

Yurov, D. V.; Prikhod'ko, V. V.

2014-11-01

The features of subcritical hybrid systems (HSs) are discussed in the context of burning up transuranic wastes from the U-Pu nuclear fuel cycle. The advantages of HSs over conventional atomic reactors are considered, and fuel cycle closure alternatives using HSs and fast neutron reactors are comparatively evaluated. The advantages and disadvantages of two HS types with neutron sources (NSs) of widely different natures -- nuclear spallation in a heavy target by protons and nuclear fusion in magnetically confined plasma -- are discussed in detail. The strengths and weaknesses of HSs are examined, and demand for them for closing the U-Pu nuclear fuel cycle is assessed.

Novel waste printed circuit board recycling process with molten salt.

PubMed

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.

Novel waste printed circuit board recycling process with molten salt

PubMed Central

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

Modelos para la Unificacion de Conceptos, Metodos y Procedimientos Administrativos (Guidelines for Uniform Administrative Concepts, Methods, and Procedures).

ERIC Educational Resources Information Center

Serrano, Jorge A., Ed.

These documents, discussed and approved during the first meeting of the university administrators affiliated with the Federation of Private Universities of Central America and Panama (FUPAC), seek to establish uniform administrative concepts, methods, and procedures, particularly with respect to budgetary matters. The documents define relevant…

La pronunciacion espanola y los metodos de investigacion. (Spanish Pronunciation and Methods of Investigation.)

ERIC Educational Resources Information Center

Torreblanca, Maximo

1988-01-01

Discusses the validity of studies of Spanish pronunciation in terms of research methods employed. Topics include data collection in the laboratory vs. in a natural setting; recorded vs. non-recorded data; quality of the recording; aural analysis vs. spectrographic analysis; and transcriber reliability. Suggestions for improving data collection are…

77 FR 38341 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-27

... Pike, Rockville, Maryland. Wednesday, July 11, 2012, Conference Room T2-B1, 11545 Rockville Pike... U.S.C. 552b(c)(4)] Thursday, July 12, 2012, Conference Room T2-B1, 11545 Rockville Pike, Rockville... Bottom,'' and (2) NUREG/CR-7040, ``Evaluation of JNES Equipment Fragility Tests for Use in Seismic...

Feasibility of Nuclear Power on U.S. Military Installations. 2nd Revision

DTIC Science & Technology

2011-03-01

Small Modular Reactor , Military Installation Energy, Energy Assurance 16. SECURITY CLASSIFICATION OF: a. REPORT I b. ABSTRACT U c. THIS PAGE i; 17. LIMITATION OF ABSTRACT SAR 18. NUMBER OF PAGES 98 19a. NAME OF RESPONSIBLE PERSON Knowledge Center/Rhea Stone 19b. TELEPHONE NUMBER (Include area code) 703-824-2110 Standard Form 298 (Rev. 8/98) Prescribed bv ANSI Sid 239.18 Contents Preliminary note: Development and commercial deployment of small modular reactors

Fusion Plasma Performance and Confinement Studies on JT-60 and JT-60U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kamada, Y.; Fujita, T.; Ishida, S.

2002-09-15

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has been developing and optimizing the operational concepts, and extending the discharge regimes toward sustainment of high integrated performance in the reactor relevant parameter regime. In addition to achievement of high fusion plasma performances such as the equivalent breakeven condition (Q{sub DT}{sup eq} up to 1.25) and a high fusion triple product n{sub D}(0){tau}{sub E}T{sub i}(0) = 1.5 x 10{sup 21} m{sup -3}skeV, JT-60U has demonstrated the integrated performance of highmore » confinement, high {beta}{sub N}, full non-inductive current drive with a large fraction of bootstrap current. These favorable performances have been achieved in the two advanced operation regimes, the reversed magnetic shear (RS) and the weak magnetic shear (high-{beta}{sub p}) ELMy H modes characterized by both internal transport barriers (ITB) and edge transport barriers (ETB). The key factors in optimizing these plasmas towards high integrated performance are control of profiles of current, pressure, rotation, etc. utilizing a variety of heating, current drive, torque input, and particle control capabilities and high triangularity operation. As represented by discovery of ITBs (density ITB in the central pellet mode, ion temperature ITB in the high-{beta}{sub p} mode, and electron temperature ITB in the reversed shear mode), confinement studies in JT-60/JT-60U have been emphasizing freedom and also restriction of radial profiles of temperature and density. In addition to characterization of confinement and analyses of transport properties of the OH, the L-mode, the H-mode, the pellet mode, the high-{beta}{sub p} mode, and the RS mode, JT-60U has clarified formation conditions, spatial structures and dynamics of edge and internal transport barriers, and evaluated effects of repetitive MHD events on confinement such as sawteeth and ELMs. Through these studies, JT-60U has demonstrated applicability of the high confinement modes to ITER and the steady-state tokamak reactors.« less

Natural and man-made radioactivity in soils and plants around the research reactor of Inshass.

PubMed

Higgy, R H; Pimpl, M

1998-12-01

The specific radioactivities of the U-series, 232Th, 137Cs and 40K were measured in soil samples around the Inshass reactor in Cairo, using a gamma-ray spectrometer with a HpGe detector. The alpha activity of 238U, 234U and 235U was measured in the same soil samples by surface barrier detectors after radiochemical separation and the obtained results were compared with the specific activities determined by gamma-measurements. The alpha-activity of 238Pu, 239+240Pu, 241Am, 242Cm and 244Cm was measured after radiochemical separation by surface barrier detectors for both soil and plant samples. Then beta-activity of 241Pu was measured using liquid scintillation spectrometry.

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

NASA Astrophysics Data System (ADS)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Modeling Microalgae Productivity in Industrial-Scale Vertical Flat Panel Photobioreactors.

PubMed

Endres, Christian H; Roth, Arne; Brück, Thomas B

2018-05-01

Potentially achievable biomass yields are a decisive performance indicator for the economic viability of mass cultivation of microalgae. In this study, a computer model has been developed and applied to estimate the productivity of microalgae for large-scale outdoor cultivation in vertical flat panel photobioreactors. Algae growth is determined based on simulations of the reactor temperature and light distribution. Site-specific weather and irradiation data are used for annual yield estimations in six climate zones. Shading and reflections between opposing panels and between panels and the ground are dynamically computed based on the reactor geometry and the position of the sun. The results indicate that thin panels (≤0.05 m) are best suited for the assumed cell density of 2 g L -1 and that reactor panels should face in north-south direction. Panel spacings of 0.4-0.75 m at a panel height of 1 m appear most suitable for commercial applications. Under these preconditions, yields of around 10 kg m -2 a -1 are possible for most locations in the U.S. Only in hot climates significantly lower yields have to be expected, as extreme reactor temperatures limit overall productivity.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 usingmore » 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.« less

Space Nuclear Facility test capability at the Baikal-1 and IGR sites Semipalatinsk-21, Kazakhstan

NASA Astrophysics Data System (ADS)

Hill, T. J.; Stanley, M. L.; Martinell, J. S.

1993-01-01

The International Space Technology Assessment Program was established 1/19/92 to take advantage of the availability of Russian space technology and hardware. DOE had two delegations visit CIS and assess its space nuclear power and propulsion technologies. The visit coincided with the Conference on Nuclear Power Engineering in Space Nuclear Rocket Engines at Semipalatinsk-21 (Kurchatov, Kazakhstan) on Sept. 22-25, 1992. Reactor facilities assessed in Semipalatinski-21 included the IVG-1 reactor (a nuclear furnace, which has been modified and now called IVG-1M), the RA reactor, and the Impulse Graphite Reactor (IGR), the CIS version of TREAT. Although the reactor facilities are being maintained satisfactorily, the support infrastructure appears to be degrading. The group assessment is based on two half-day tours of the Baikals-1 test facility and a brief (2 hr) tour of IGR; because of limited time and the large size of the tour group, it was impossible to obtain answers to all prepared questions. Potential benefit is that CIS fuels and facilities may permit USA to conduct a lower priced space nuclear propulsion program while achieving higher performance capability faster, and immediate access to test facilities that cannot be available in this country for 5 years. Information needs to be obtained about available data acquisition capability, accuracy, frequency response, and number of channels. Potential areas of interest with broad application in the U.S. nuclear industry are listed.

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

DOE PAGES

Kim, Ki-Hwan; Kim, Jong-Hwan; Oh, Seok-Jin; ...

2016-01-01

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the developmentmore » of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.« less

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, W. E.; Rudisill, T. S.; O'Rourke, P. E.

2017-01-26

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgasmore » composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.« less

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.

2015-12-15

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket basedmore » on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.« less

The Nuclear Renaissance in the U.S.

ScienceCinema

Buongiorno, Jacopo

2018-04-23

Nuclear power currently provides 20% of the electricity generation in the U.S. and about 16% worldwide.  As a carbon-free energy source, nuclear is receiving a lot of attention by industry, lawmakers and environmental groups, as they attempt to resolve the issue of man-made climate change.  For the first time in 30 years several U.S. electric utilities have applied for construction and operation licenses of new nuclear power plants.  This talk will review the safety, operational and economic record of the existing U.S. commercial reactor fleet, will provide an overview of the reactor designs considered for the new wave of plant construction, and will discuss several research projects being conducted at the Massachusetts Institute of Technology to support the expansion of nuclear power in the U.S. and overseas.

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

NASA Astrophysics Data System (ADS)

Nevinitsa, V. A.; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

2015-12-01

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing 233U from 232Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

Transuranic inventory reduction in repository by partitioning and transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kang, C.H.; Kazimi, M.S.

1992-01-01

The promise of a new reprocessing technology and the issuance of Environmental Protection Agency (EPA) and U.S. Nuclear Regulatory Commission regulations concerning a geologic repository rekindle the interest in partitioning and transmutation of transuranic (TRU) elements from discharged reactor fuel as a high level waste management option. This paper investigates the TRU repository inventory reduction capability of the proposed advanced liquid metal reactors (ALMRs) and integral fast reactors (IFRs) as well as the plutonium recycled light water reactors (LWRs).

Determining Pu-239 content by resonance transmission analysis using a filtered reactor beam.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klann, R. T.

A novel technique has been developed at Argonne National Laboratory to determine the {sup 239}Pu content in EBR-II blanket elements using resonance transmission analysis (RTA) with a filtered reactor beam. The technique uses cadmium and gadolinium filters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range from 0.1 to 0.5 eV, the total microscopic cross-section of {sup 239}Pu is significantly larger than the cross-sections of {sup 238}U and {sup 235}U. This large difference in cross-section allows small amounts of {sup 239}Pu to be detected in uranium samples. Tests usingmore » a direct beam from a 250 kW TRIGA reactor have been performed with stacks of depleted uranium and {sup 239}Pu foils. Preliminary measurement results are in good agreement with the predicted results up to about two weight percent of {sup 239}Pu in the sample. In addition, measured {sup 239}Pu masses were in agreement with actual sample masses with uncertainties less than 3.8 percent.« less

77 FR 58420 - Advisory Committee On Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-20

... Pike, Rockville, Maryland. Thursday, October 4, 2012, Conference Room T2-B1, 11545 Rockville Pike....: Safety Evaluation Report (SER) Associated with WCAP-16793-NP, Revision 2, ``Evaluation of Long-Term..., ``Evaluation of JNES Equipment Fragility Tests for Use in Seismic Probabilistic Risk Assessments for U.S...

75 FR 12804 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR; Notice...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-17

... meeting on April 6, 2010, at 11545 Rockville Pike, T2-B1, Rockville, Maryland. The entire meeting will be... Chapters 11 and 16 of the Safety Evaluation Report with Open Items associated with the U.S. EPR Design...

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period hasmore » been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.« less

Extension of the TRANSURANUS burnup model to heavy water reactor conditions

NASA Astrophysics Data System (ADS)

Lassmann, K.; Walker, C. T.; van de Laar, J.

1998-06-01

The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After verification against data for the radial distributions of Xe, Cs, Nd and Pu from electron probe microanalysis, the model is used to analyse the formation of the high burnup structure in a heavy water reactor. The new model allows the analysis of light water reactor fuel rod designs at high burnup in the OECD Halden Heavy Water Reactor.

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

Short communication on " In-situ TEM ion irradiation investigations on U 3Si 2 at LWR temperatures"

DOE PAGES

Miao, Yinbin; Harp, Jason; Mo, Kun; ...

2016-11-21

Here, the radiation-induced amorphization of U 3Si 2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 10 15 ions/cm 2 to examine their amorphization behavior under light water reactor (LWR) conditions. U 3Si 2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"

NASA Astrophysics Data System (ADS)

Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

2017-02-01

The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

U.S. Proliferation Policy and the Campaign Against Transnational Terror: Linking the U.S. Non-Proliferation Regime to Homeland Security Efforts

DTIC Science & Technology

2013-12-01

device (IED) in a public park is more difficult than setting off an IED in a secured government building . Alternatively, constructing a pipe bomb is...against nuclear and power industry installations” intended to “seize nuclear materials and use them to build WMD for their own political use.” 5...reactor under construction .469 Damascus faces unresolved allegations that it illicitly tried to build a plutonium production reactor at a site

Redefining Success: Applying Lessons in Nuclear Diplomacy from North Korea to Iran (INSS Strategic Perspectives, Number 1, September 2010)

DTIC Science & Technology

2010-09-01

DISTRIBUTION/AVAILABILITY STATEMENT Approved for public release; distribution unlimited 13 . SUPPLEMENTARY NOTES 14. ABSTRACT 15. SUBJECT TERMS...consideration of a U.S. military strike), peacefully return North Korea to the NPT and freeze its nuclear program 19 months later. 13 In the process...lifted the nuclear freeze and in January 2003 withdrew from 13 Redefining Success the NPT, citing two main reasons: U.S. failure to honor the reactor

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

NASA Astrophysics Data System (ADS)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite-reflected cylindrical geometry.

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-10-28

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008; and, Chris...related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-12-23

reactors deployed” in the UAE. Some Members of Congress had welcomed the UAE government’s stated commitments not to pursue proliferation-sensitive...for the planned nuclear reactor or on handling spent reactor fuel. (...continued) May...firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

10 CFR 51.121 - Status of NEPA actions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... address inquiries to: (a) Utilization facilities: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission... NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND...

10 CFR 51.121 - Status of NEPA actions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... address inquiries to: (a) Utilization facilities: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission... NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND...

10 CFR 51.121 - Status of NEPA actions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... address inquiries to: (a) Utilization facilities: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission... NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND...

Contribution of extracellular polymeric substances from Shewanella sp. HRCR-1 biofilms to U(VI) immobilization.

PubMed

Cao, Bin; Ahmed, Bulbul; Kennedy, David W; Wang, Zheming; Shi, Liang; Marshall, Matthew J; Fredrickson, Jim K; Isern, Nancy G; Majors, Paul D; Beyenal, Haluk

2011-07-01

The goal of this study was to quantify the contribution of extracellular polymeric substances (EPS) to U(VI) immobilization by Shewanella sp. HRCR-1. Through comparison of U(VI) immobilization using cells with bound EPS (bEPS) and cells with minimal EPS, we show that (i) bEPS from Shewanella sp. HRCR-1 biofilms contribute significantly to U(VI) immobilization, especially at low initial U(VI) concentrations, through both sorption and reduction; (ii) bEPS can be considered a functional extension of the cells for U(VI) immobilization and they likely play more important roles at lower initial U(VI) concentrations; and (iii) the U(VI) reduction efficiency is dependent upon the initial U(VI) concentration and decreases at lower concentrations. To quantify the relative contributions of sorption and reduction to U(VI) immobilization by EPS fractions, we isolated loosely associated EPS (laEPS) and bEPS from Shewanella sp. HRCR-1 biofilms grown in a hollow fiber membrane biofilm reactor and tested their reactivity with U(VI). We found that, when reduced, the isolated cell-free EPS fractions could reduce U(VI). Polysaccharides in the EPS likely contributed to U(VI) sorption and dominated the reactivity of laEPS, while redox active components (e.g., outer membrane c-type cytochromes), especially in bEPS, possibly facilitated U(VI) reduction.

Contribution of Extracellular Polymeric Substances from Shewanella sp. HRCR-1 Biofilms to U(VI) Immobilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Bin; Ahmed, B.; Kennedy, David W.

2011-06-05

The goal of this study was to quantify the contribution of extracellular polymeric substances (EPS) in U(VI) immobilization by Shewanella sp. HRCR-1. Through comparison of U(VI) immobilization using cells with bound EPS (bEPS) and cells without EPS, we showed that i) bEPS from Shewanella sp. HRCR-1 biofilms contributed significantly to U(VI) immobilization, especially at low initial U(VI) concentrations, through both sorption and reduction; ii) bEPS could be considered as a functional extension of the cells for U(VI) immobilization and they likely play more important roles at initial U(VI) concentrations; and iii) U(VI) reduction efficiency was found to be dependent uponmore » initial U(VI) concentration and the efficiency decreased at lower concentrations. To quantify relative contribution of sorption and reduction in U(VI) immobilization by EPS fractions, we isolated loosely associated EPS (laEPS) and bEPS from Shewanella sp. HRCR-1 biofilms grown in a hollow fiber membrane biofilm reactor and tested their reactivity with U(V). We found that, when in reduced form, the isolated cell-free EPS fractions could reduce U(VI). Polysaccharides in the EPS likely contributed to U(VI) sorption and dominated reactivity of laEPS while redox active components (e.g., outer membrane c-type cytochromes), especially in bEPS, might facilitate U(VI) reduction.« less

Microstructural characterization of a thin film ZrN diffusion barrier in an As-fabricated U-7Mo/Al matrix dispersion fuel plate

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Perez, Emmanuel; Wiencek, Tom; Leenaers, Ann; Van den Berghe, Sven

2015-03-01

The United States High Performance Research Reactor Fuel Development program is developing low enriched uranium fuels for application in research and test reactors. One concept utilizes U-7 wt.% Mo (U-7Mo) fuel particles dispersed in Al matrix, where the fuel particles are coated with a 1 μm-thick ZrN coating. The ZrN serves as a diffusion barrier to eliminate a deleterious reaction that can occur between U-7Mo and Al when a dispersion fuel is irradiated under aggressive reactor conditions. To investigate the final microstructure of a physically-vapor-deposited ZrN coating in a dispersion fuel plate after it was fabricated using a rolling process, characterization samples were taken from a fuel plate that was fabricated at 500 °C using ZrN-coated U-7Mo particles, Al matrix and AA6061 cladding. Scanning electron and transmission electron microscopy analysis were performed. Data from these analyses will be used to support future microstructural examinations of irradiated fuel plates, in terms of understanding the effects of irradiation on the ZrN microstructure, and to determine the role of diffusion barrier microstructure in eliminating fuel/matrix interactions during irradiation. The as-fabricated coating was determined to be cubic-ZrN (cF8) phase. It exhibited a columnar microstructure comprised of nanometer-sized grains and a region of relatively high porosity, mainly near the Al matrix. Small impurity-containing phases were observed at the U-7Mo/ZrN interface, and no interaction zone was observed at the ZrN/Al interface. The bonding between the U-7Mo and ZrN appeared to be mechanical in nature. A relatively high level of oxygen was observed in the ZrN coating, extending from the Al matrix in the ZrN coating in decreasing concentration. The above microstructural characteristics are discussed in terms of what may be most optimal for a diffusion barrier in a dispersion fuel plate application.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lane, Taylor; Parma, Edward J.

Delayed fission gamma-rays play an important role in determining the time dependent ioniz- ing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from acti- vation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is nec- essary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution tomore » the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray character- istics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 [ 1 ] and experimental data [ 2 , 3 ] and other published literature, includ- ing ENDF/B-VII.1 [ 4 ]. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.« less

76 FR 52716 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-23

... Register on Thursday, October 21, 2010 (75 FR 65038-65039). Thursday, September 8, 2011, Conference Room T2... low-level waste disposal and site- specific analysis. 1:15 p.m.-3:15 p.m.: Safety Evaluation Report... U.S.C. 552b(c)(4).] Friday, September 9, 2011, Conference Room T2-B1, 11545 Rockville Pike...

76 FR 16457 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-23

... Thursday, October 21, 2010 (74 FR 65038-65039). Thursday, April 7, 2011, Conference Room T2-B1, 11545....: Selected Chapters of the Safety Evaluation Report (SER) with Open Items Associated with the Calvert Cliffs... pursuant to 5 U.S.C. 552b(c)(4).] Friday, April 8, 2011, Conference Room T2-B1, 11545 Rockville Pike...

75 FR 13799 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-23

..., 2009, (74 FR 52829-52830). Thursday, April 8, 2010, Conference Room T2-B1, Two White Flint North... Chapters of the Safety Evaluation Report (SER) with Open Items Associated with the Review of the U.S..., Conference Room T2-B1, Two White Flint North, Rockville, Maryland 8:30 a.m.-8:35 a.m.: Opening Remarks by the...

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. Wigeland; T. Taiwo; M. Todosow

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there wasmore » no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.« less

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

NASA Astrophysics Data System (ADS)

Rouf; Su'ud, Zaki

2016-08-01

Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

Continuous Production of Ethanol from Starch Using Glucoamylase and Yeast Co-Immobilized in Pectin Gel

NASA Astrophysics Data System (ADS)

Giordano, Raquel L. C.; Trovati, Joubert; Schmidell, Willibaldo

This work presents a continuous simultaneous saccharification and fermentation (SSF) process to produce ethanol from starch using glucoamylase and Saccharomyces cerevisiae co-immobilized in pectin gel. The enzyme was immobilized on macroporous silica, after silanization and activation of the support with glutaraldehyde. The silicaenzyme derivative was co-immobilized with yeast in pectin gel. This biocatalyst was used to produce ethanol from liquefied manioc root flour syrup, in three fixed bed reactors. The initial reactor yeast load was 0.05 g wet yeast/ml of reactor (0.1 g wet yeast/g gel), used in all SSF experiments. The enzyme concentration in the reactor was defined by running SSF batch assays, using different amount of silica-enzyme derivative, co-immobilized with yeast in pectin gel. The chosen reactor enzyme concentration, 3.77 U/ml, allowed fermentation to be the rate-limiting step in the batch experiment. In this condition, using initial substrate concentration of 166.0 g/1 of total reducing sugars (TRS), 1 ml gel/1 ml of medium, ethanol productivity of 8.3 g/l/h was achieved, for total conversion of starch to ethanol and 91% of the theoretical yield. In the continuous runs, feeding 163.0 g/1 of TRS and using the same enzyme and yeast concentrations used in the batch run, ethanol productivity was 5.9 g ethanol/1/h, with 97% of substrate conversion and 81% of the ethanol theoretical yield. Diffusion effects in the extra-biocatalyst film seemed to be reduced when operating at superficial velocities above 3.7 × 10-4 cm/s.

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

NASA Astrophysics Data System (ADS)

Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

2015-12-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yun, D.; Taiwo, T. A.; Kim, T. K.

2010-10-01

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

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety ofmore » LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.« less

76 FR 65541 - Assuring the Availability of Funds for Decommissioning Nuclear Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-21

... NUCLEAR REGULATORY COMMISSION [NRC-2009-0263] Assuring the Availability of Funds for Decommissioning Nuclear Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC or Commission) is issuing a revision to Regulatory...

An experimental and modeling study of grain-scale uranium desorption from field-contaminated sediments and the potential influence of microporosity on mass-transfer

NASA Astrophysics Data System (ADS)

Stoliker, D.; Liu, C.; Kent, D. B.; Zachara, J. M.

2012-12-01

The aquifer below the 300-Area of the Hanford site (Richland, WA, USA) is plagued by a persistent plume of dissolved uranium (U(VI)) in excess of the Environmental Protection Agency drinking water maximum contamination level even after the removal of highly contaminated sediments. The aquifer sediments in the seasonally saturated lower vadose zone act as both a source and sink for uranium during stage changes in the nearby Columbia River. Diffusion limitation of uranium mass-transfer within these sediments has been cited as a potential cause of the plume's persistence. Equilibrium U(VI) sorption is a strong function of variable chemical conditions, especially carbonate, hydrogen, and uranyl ion activities. Field-contaminated sediments from the site require up to 1,000 hours to reach equilibrium in static batch reactors. Increases in U(VI) concentrations over longer time-scales result from changes in chemical conditions, which drive reactions with sediments that favor U(VI) desorption. Grain-scale U(VI) sorption/desorption rates are slow, likely owing to diffusion of U(VI) and other solutes through intra-granular pore domains. In order to improve understanding of the impact of intra-granular diffusion and chemical reactions controlling grain-scale U(VI) release, experiments were conducted on individual particle size fractions of a <8 mm composite of field-contaminated, lower vadose zone sediments. For each size fraction, equilibrium U(VI) sorption/desorption in static batch reactors was well-described by surface complexation models over a range of chemical conditions applicable to the field site. Desorption rates from individual size fractions in flow-through batch reactors, examined under a single set of constant chemical conditions with multiple stop-flow events, were similar for all size fractions <2 mm. Kinetic U(VI) desorption in flow-through batch reactors was modeled using a multi-rate surface complexation approach, where sorption/desorption rates were assumed to be proportional to the displacement from equilibrium and multiple diffusion domains were described with a two-parameter lognormal distribution of mass-transfer rate coefficients. Parameters describing mass transfer were the same for all size fractions <2 mm but differed for the largest (2-8 mm) size fraction. The evolution of pH, along with dissolved cation and carbonate concentrations, was modeled using equilibrium cation exchange, rate-limited calcite dissolution, aerobic respiration, and silica dissolution. Desorption and chemical reaction models calibrated with individual size fractions predicted U(VI) and chemical composition as a function of time for the bulk sediment sample. Volumes of pores less than 2.4 nm, quantified using nitrogen adsorption-desorption isotherms, were the same for all size fractions < 2 mm, nearly double that of the 2-8 mm size fraction. Similarity in the observed pore volumes and multi-rate mass-transfer parameters across all size fractions <2 mm suggest the importance of pores in this size class in controlling slow grain-scale U(VI) desorption rates. Models like these provide a means for testing the influence of grain-scale mass-transfer on the persistence of U(VI) plume at the site.

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

The use of MOX caramel fuel mixed with 241Am, 242mAm and 243Am as burnable absorber actinides for the MTR research reactors.

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2017-07-01

The MOX (UO 2 &PuO 2 ) caramel fuel mixed with 241 Am, 242m Am and 243 Am as burnable absorber actinides was proposed as a fuel of the MTR-22MW reactor. The MCNP4C code was used to simulate the MTR-22MW reactor and estimate the criticality and the neutronic parameters, and the power peaking factors before and after replacing its original fuel (U 3 O 8 -Al) by the MOX caramel fuel mixed with 241 Am, 242m Am and 243 Am actinides. The obtained results of the criticality, the neutronic parameters, and the power peaking factors for the MOX caramel fuel mixed with 241 Am, 242m Am and 243 Am actinides were compared with the same parameters of the U 3 O 8 -Al original fuel and a maximum difference is -6.18% was found. Additionally, by recycling 2.65% and 2.71% plutonium and 241 Am, 242m Am and 243 Am actinides in the MTR-22MW reactor, the level of 235 U enrichment is reduced from 4.48% to 3% and 2.8%, respectively. This also results in the reduction of the 235 U loading by 32.75% and 37.22% for the 2.65%, the 2.71% plutonium and 241 Am, 242m Am and 243 Am actinides, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

NASA Astrophysics Data System (ADS)

Kochetkov, Anatoly; Krása, Antonín; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente; Bianchini, Giancarlo; Fabrizio, Valentina; Carta, Mario; Firpo, Gabriele; Fridman, Emil; Sarotto, Massimo

2017-09-01

During the GUINEVERE FP6 European project (2006-2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

Development of a prototype infrared imaging bolometer for NSTX-U

NASA Astrophysics Data System (ADS)

van Eden, G. G.; Delgado-Aparicio, L. F.; Gray, T. K.; Jaworski, M. A.; Morgan, T. W.; Peterson, B. J.; Reinke, M. L.; Sano, R.; Mukai, K.; Differ/Pppl Collaboration; Nifs/Pppl Collaboration

2015-11-01

Measurements of the radiated power in fusion reactors are of high importance for studying detachment and the overall power balance. A prototype Infrared Video Bolometer (IRVB) is being developed for NSTX-U complementing resistive bolometer and AXUV diode diagnostics. The IRVB has proven to be a powerful tool on LHD and JT-60U for its 2D imaging quality and reactor environment compatibility. For NSTX-U, a poloidal view of the lower center stack and lower divertor are envisaged for the 2016 run campaign. The IRVB concept images radiation from the plasma onto a 2.5 μm thick 9 x 7 cm2 calibrated Pt foil and monitors its temperature evolution using an IR camera (SB focal plane, 2-12 μm, 128x128 pixels, 1.6 kHz). The power incident on the foil is calculated by solving the 2D +time heat diffusion equation. Benchtop characterization is presented, demonstrating a sensitivity of approximately 20 mK and a noise equivalent power density of 71.5 μW cm-2 for 4x20 bolometer super-pixels and a 50 Hz time response. The hardware design, optimization of camera and detector settings as well as first results of both synthetic and experimental origin are discussed.

CHI Research on NSTX-U

NASA Astrophysics Data System (ADS)

Lay, W.-S.; Raman, R.; Jarboe, T. R.; Nelson, B. A.; Mueller, D.; Ebrahimi, F.; Ono, M.; Jardin, S. C.; Taylor, G.

2017-10-01

At present about 20% of the total plasma current required for sustained operation has been generated by transient CHI. The present understanding suggests that it may be possible to generate all of the needed current in a ST / tokamak using transient CHI. In such a scenario, one could transition directly from a CHI produced plasma to a non-inductively sustained plasma, without the difficult intermediate step that involves non-inductive current ramp-up. STs based on this new configuration would take advantage of evolving developments in high-temperature superconductor technology to develop a simpler design ST that relies primarily on CHI for plasma current generation. Motivated by the very good results from NSTX and HIT-II, we are examining the potential application of transient CHI for reactor configurations through these studies. (1) Study of the maximum levels of start-up currents that could be generated on NSTX-U, (2) application of a single biased electrode configuration on QUEST to protect the insulator from neutron damage in a CHI reactor installation, and (3) QUEST-like, but a double biased electrode configuration for PEGASUS and NSTX-U. Results from these on-going studies will be described. This work is supported by U.S. DOE Contracts: DE-AC02-09CH11466, DE-FG02-99ER54519 AM08, and DE-SC0006757.

Cultivation of E. coli in single- and ten-stage tower-loop reactors.

PubMed

Adler, I; Schügerl, K

1983-02-01

E. Coli was cultivated in batch and continuous operations in the presence of an antifoam agent in stirred-tank and in single- and ten-stage airlift tower reactors with an outer loop. The maximum specific growth rate, mu(m), the substrate yield coefficient, Y(x/s), the respiratory quotient, RQ, substrate conversion, U(s), the volumetric mass transfer coefficient, K(L)a, the specific interfacial area, a, and the specific power input, P/V(L), were measured and compared. If a medium is used with a concentration of complex substrates (extracts) 2.5 times higher than that of glucose, a spectrum of C sources is available and cell regulation influences reactor performance. Both mu(m) and Y(X/S), which were evaluated in batch reactors, cannot be used for continuous reactors or, when measured in stirred-tank reactors, cannot be employed for tower-loop reactors: mu(m) is higher in the stirred-tank batch than in the tower-loop batch reactor, mu(m) and Y(x/s) are higher in the continuous reactor than in the batch single-stage tower-loop reactor. The performance of the single-stage is better than that of the ten-stage reactor due to the inefficient trays employed. A reduction of the medium recirculation rate reduces OTR, U(s), Pr, and Y(X/S) and causes cell sedimentation and flocculation. The volumetric mass transfer coefficient is reduced with increasing cultivation time; the Sauter bubble diameter, d(s), remains constant and does not depend on operational conditions. An increase in the medium recirculation rate reduces k(L)a. The specific power input, P/V(L), for the single-stage tower loop is much lower with the same k(L)a value than for a stirred tank. The relationship k(L)a vs. P/V(L) evaluated for model media in stirred tanks, can also be used for cultivations in these reactors.

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

PubMed

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

On the Use of a Molten Salt Fast Reactor to Apply an Idealized Transmutation Scenario for the Nuclear Phase Out

PubMed Central

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described. PMID:24690768

The Use of Thorium within the Nuclear Power Industry - 13472

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Keith

2013-07-01

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ∼0.7% fissile {sup 235}U isotope, natural thorium does not contain any fissile material and is made up of the fertile {sup 232}Th isotope only. Therefore thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with fissile {sup 235}U or {sup 239}Pu in nuclear research and power reactors for conversion to fissile {sup 233}U, thereby enlarging fissile material resources. During the pioneering years of nuclear energy, frommore » the mid 1950's to mid 1970's, there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. The initial enthusiasm for thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, and improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles. (authors)« less

U.S.-Russian Civilian Nuclear Cooperation Agreement: Issues for Congress

DTIC Science & Technology

2010-07-09

for nuclear cooperation in 1973 to allow for cooperation in controlled thermonuclear fusion, fast breeder reactors , and fundamental research. The...that a 123 agreement is needed to implement this action plan—for example, full scale technical cooperation on fast reactors and demonstration of...superpowers convened a Joint Coordinating Committee for Civilian Reactor Safety starting in 1988.10 After the fall of the Soviet Union and prior to July

Renewability and sustainability aspects of nuclear energy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Şahin, Sümer, E-mail: ssahin@atilim.edit.tr

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2};more » 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ∼1.3.« less

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.

2015-12-15

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

77 FR 67837 - Callaway Plant, Unit 1; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-14

... methods: Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012... search, select ``ADAMS Public Documents'' and then select ``Begin Web- based ADAMS Search.'' For problems... INFORMATION CONTACT: Carl F. Lyon, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear...

METHOD OF PRODUCING U$sup 23$$sup 3$

DOEpatents

Seaborg, G.T.; Stoughton, R.W.

1960-08-30

A method for producing U/sup 233/ is outlined in which a body of thorium carbonate is heated to at least 200 deg C until it attains a constant weight and compressing the body into a pellet having a density of at least 2.6 g/cm/sup 3/. The pellet is enclosed in a sealed container and placed in the blanket of a thermal nuclear reactor having a neutron flux in which the majority of neutrons have an energy of below I Mev. The pellet is removed from the flux before the ratio of U/sup 233/ to Th/sup 232/ is about 1: 100.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2008-06-01

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 fuelmore » 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.« less

The effect of silicon on the interaction between metallic uranium and aluminum: A 50 year long diffusion experiment

NASA Astrophysics Data System (ADS)

Leenaers, A.; Detavernier, C.; Van den Berghe, S.

2008-11-01

The core of the BR1 research reactor at SCK•CEN, Mol (Belgium) has a graphite matrix loaded with fuel rods consisting of a natural uranium slug in aluminum cladding. The BR1 reactor has been in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, some of the fuel rods have been examined. Fabrication reports indicate that a so-called AlSi bonding layer and an U(Al,Si) 3 anti-diffusion layer on the natural uranium fuel slug were applied to limit the interaction between the uranium fuel and aluminum cladding. The microstructure of the fuel, bonding and anti-diffusion layer and cladding were analysed using optical microscopy, scanning electron microscopy and electron microprobe analysis. It was found that the AlSi bonding layer does provide a tight bond between fuel and cladding but that it is a thin USi layer that acts as effective anti-diffusion layer and not the intended U(Al,Si) 3 layer.

USHPRR FUEL FABRICATION PILLAR: FABRICATION STATUS, PROCESS OPTIMIZATIONS, AND FUTURE PLANS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wight, Jared M.; Joshi, Vineet V.; Lavender, Curt A.

The Fuel Fabrication (FF) Pillar, a project within the U.S. High Performance Research Reactor Conversion program of the National Nuclear Security Administration’s Office of Material Management and Minimization, is tasked with the scale-up and commercialization of high-density monolithic U-Mo fuel for the conversion of appropriate research reactors to use of low-enriched fuel. The FF Pillar has made significant steps to demonstrate and optimize the baseline co-rolling process using commercial-scale equipment at both the Y-12 National Security Complex (Y-12) and BWX Technologies (BWXT). These demonstrations include the fabrication of the next irradiation experiment, Mini-Plate 1 (MP-1), and casting optimizations at Y-12.more » The FF Pillar uses a detailed process flow diagram to identify potential gaps in processing knowledge or demonstration, which helps direct the strategic research agenda of the FF Pillar. This paper describes the significant progress made toward understanding the fuel characteristics, and models developed to make informed decisions, increase process yield, and decrease lifecycle waste and costs.« less

Advanced In-Pile Instrumentation for Materials Testing Reactors

NASA Astrophysics Data System (ADS)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

2014-08-01

The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research 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, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

Design Analysis of a Prepackaged Nuclear Power Plant for an Ice Cap Location

DTIC Science & Technology

1959-01-15

requirements and heating load 1.3 Site Conditions 1,U Air Transportability 1.5 Standby Power Availability 1.6 Building Structuree and Foundations 2,0...Skid with Reactor and Steam Generator Generator Weight Distribution Foundation Load Diagram (Secondary) Turbine Generator Package - Typical...Requirements and Heating Load The plant shall be capable of producing a minimum of 1500 Kw net ^ electrical energy at 4160/2400 volts, three phase

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D. -S.

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data setmore » of full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model.« less

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

A SPACESHIP WITH NUCLEAR PROPULSION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Polorny, J.

1962-01-01

ABS>A proposed space vehicle with nuclear propulsion for a round-trip Martian mission is described. It would be powered by a 270-Mw graphite- moderated, U-fueled nuclear reactor with a core 1 m high by 1 m in diameter, and use gas as propellant. The gas would be heated to the maximum temperature in the reactor and additionally accelerated by an electromagnetic field. To this end, small quantities of K would be injected into the gas stream to increase its electric conductivity. The required electrical energy would be produced by liquid-Na-cooled thermionic converters. The vehicle would weigh 115000 kg, including 43000 kgmore » of H propellant with tankage, and 7000 kg of sustenance material for one year. Chemical rockets would launch the vehicle with a crew of three men into an earth orbit where nuclear propulsion would take over. Upon reactor start-up, three heat exchangers (minimum dimensions 30 x 18 m) would be fanned out. A shielded well with a diameter of 2.5 m would protect the crew from radiation during reactor operation, passage through the earth radiation belts, and at periods of solar flares. (OTS)« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Thermophysical properties of U 3 Si 2 to 1773K

DOE PAGES

White, Joshua Taylor; Nelson, Andrew Thomas; Dunwoody, John Tyler; ...

2015-05-08

Use of U 3Si 2 in nuclear reactors requires accurate thermophysical property data to capture heat transfer within the core. Compilation of the limited previous research efforts focused on the most critical property, thermal conductivity, reveals extensive disagreement. Assessment of this data is challenged by the fact that the critical structural and chemical details of the material used to provide historic data is either absent or confirms the presence of significant impurity phases. This study was initiated to fabricate high purity U 3Si 2 to quantify the coefficient of thermal expansion, heat capacity, thermal diffusivity, and thermal conductivity from roommore » temperature to 1773 K. Here, the datasets provided in this manuscript will facilitate more detailed fuel performance modeling to assess both current and proposed reactor designs that incorporate U 3Si 2.« less

Eddy Current Flow Measurements in the FFTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

Feasibility of rotating fluidized bed reactor for rocket propulsion

NASA Technical Reports Server (NTRS)

Ludewig, H.; Manning, A. J.; Raseman, C. J.

1974-01-01

The rotating fluidized bed reactor concept is outlined, and its application to rocket propulsion is discussed. Experimental results obtained indicate that minimum fluidization correlations commonly in use for 1-g beds can also be applied to multiple-g beds. It was found that for a low thrust system (20,000 lbf) the fuel particle size and/or particle stress play a limiting role on performance. The superiority of U-233 as a fuel for this type of rocket engine is clearly demonstrated in the analysis. The maximum thrust/weight ratio for a 90,000N thrust engine was found to be approximately 65N/kg.

Silica-Immobilized Enzyme Reactors; Application to Cholinesterase-Inhibition Studies

DTIC Science & Technology

2006-03-01

a i a i e v b c a m p l 1 d Journal of Chromatography B, 843 (2006) 310–316 Silica-immobilized enzyme reactors; application to...however, have specific omat d t t i w i h c u C p R n t o t fl p f o l p d v t o s fl t t c t a b f c i h a t s m i b 2 2 ≈ c s 0 c A R I H l 2 b p a y ε c...b r T p d ( m 2 w

El Metodo Llamado Proyecto (The Project Approach). ERIC Digest.

ERIC Educational Resources Information Center

Katz, Lilian G.

A project is an in-depth investigation of a topic worth learning more about, usually undertaken by a group of children within a class. The goal of a project is to learn more about a topic rather than to find answers to questions posed by a teacher. Project work is complementary to the systematic parts of a curriculum. Whereas systematic…

"Mastery Learning" Como Metodo Psicoeducativo para Ninos con Problemas Especificos de Aprendizaje. ("Mastery Learning" as a Psychoeducational Method for Children with Specific Learning Problems.)

ERIC Educational Resources Information Center

Coya, Liliam de Barbosa; Perez-Coffie, Jorge

1982-01-01

"Mastery Learning" was compared with the "conventional" method of teaching reading skills to Puerto Rican children with specific learning disabilities. The "Mastery Learning" group showed significant gains in the cognitive and affective domains. Results suggested Mastery Learning is a more effective method of teaching…

Exploding the myths about the fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burns, S.

1979-01-01

This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

Solar Power Satellites - A Review of the Space Transportation Options.

DTIC Science & Technology

1980-03-01

already exists with such systems, gained mainly through liquid-metal breeder reactor programmes. 0 For example, inlet temperatures of 970 C can be handled...alternatives exist. In addition, there would be extreme reluctance on the part of most governments to allow large C- reactors , producing gigawatts of power, to...antenna. The reactors employed are high-temperature gas- cooled breeders , which convert U238 into fissile plutonium. Each of the modules includes a

Japan’s Nuclear Future: Policy Debate, Prospects, and U.S. Interests

DTIC Science & Technology

2008-05-09

raised in particular over the construction of an industrial- scale reprocessing facility in Japan,. Additionally, fast breeder reactors also produce more...Nuclear Fuel Cycle Engineering Laboratories. 10 A fast breeder reactor is a fast neutron reactor that produces more plutonium than it consumes, which can...Japan Nuclear Fuel Limited (JNFL) has built and is currently running active testing on a large - scale commercial reprocessing plant at Rokkasho-mura

AQUABOX 50 AND MARABU PACKED BIOLOGICAL REACTOR SYSTEM TECHNOLOGY EVALUATION, STADTWERKE DUESSELDORF AG SITE, DUESSELDORF, GERMANY

EPA Science Inventory

This ITER summarizes the results of an evaluation of the AQUABOX 50 and MARABU Packed Biological Reactor technologies. The evaluation was conducted under a bilateral agreement between the United States (U.S.) Environmental Protection Agency (EPA) Superfund Innovative Technology ...

U.S. Nuclear Cooperation With India: Issues for Congress

DTIC Science & Technology

2010-02-24

Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor Export Deal,” Chennai Business Line Online, July 10, 2009...agreements that covered reactors producing more than 5 MW thermal or special nuclear material connected therewith. 123 United States General Accounting

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR... under them should be sent by mail addressed to: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission...

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Communications. 55.5 Section 55.5 Energy NUCLEAR...) By mail addressed to—Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; or (2) By delivery...

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Communications. 55.5 Section 55.5 Energy NUCLEAR...) By mail addressed to—Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; or (2) By delivery...

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Matos, J.E.

At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U{sub 3}Si{sub 2} fuel is about 6.0 g U/cm{sup 3}. The French Commissariat a l`Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L`Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm{sup 3} and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersionmore » fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm{sup 3}. On the other hand, UN is the least reactive fuel because of the relatively large {sup 14}N(n,p) cross section. For a fixed value of k{sub eff}, the required {sup 235}U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO{sub 2} dispersions are only useful for uranium densities below 5.0 g/cm{sup 3}. In this density range, however, UO{sub 2} is more reactive than U{sub 3}Si{sub 2}.« less

Analysis of the key enzymes of butyric and acetic acid fermentation in biogas reactors

PubMed Central

Gabris, Christina; Bengelsdorf, Frank R; Dürre, Peter

2015-01-01

This study aimed at the investigation of the mechanisms of acidogenesis, which is a key process during anaerobic digestion. To expose possible bottlenecks, specific activities of the key enzymes of acidification, such as acetate kinase (Ack, 0.23–0.99 U mg−1 protein), butyrate kinase (Buk, < 0.03 U mg−1 protein) and butyryl-CoA:acetate-CoA transferase (But, 3.24–7.64 U mg−1 protein), were determined in cell free extracts of biogas reactor content from three different biogas reactors. Furthermore, the detection of Ack was successful via Western blot analysis. Quantification of corresponding functional genes encoding Buk (buk) and But (but) was not feasible, although an amplification was possible. Thus, phylogenetic trees were constructed based on respective gene fragments. Four new clades of possible butyrate-producing bacteria were postulated, as well as bacteria of the genera Roseburia or Clostridium identified. The low Buk activity was in contrast to the high specific But activity in the analysed samples. Butyrate formation via Buk activity does barely occur in the investigated biogas reactor. Specific enzyme activities (Ack, Buk and But) in samples drawn from three different biogas reactors correlated with ammonia and ammonium concentrations (NH3 and NH4+-N), and a negative dependency can be postulated. Thus, high concentrations of NH3 and NH4+-N may lead to a bottleneck in acidogenesis due to decreased specific acidogenic enzyme activities. PMID:26086956

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

DOEpatents

Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

1982-03-31

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

DOEpatents

Caldwell, John T.; Kunz, Walter E.; Atencio, James D.

1984-01-01

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

Dissecting Reactor Antineutrino Flux Calculations

NASA Astrophysics Data System (ADS)

Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

2017-09-01

Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

Dissecting Reactor Antineutrino Flux Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

2017-09-15

Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235 U , 239 Pu , 241 Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In our present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238 U contribution as wellmore » as the effective charge and the allowed shape assumption used in the conversion method. Here, we observe that including a shape correction of about + 6 % MeV - 1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.« less

Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Lahoda, E.; Wenner, M.

2013-07-01

This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principlemore » be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation atmore » inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.« less

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE PAGES

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

2015-06-21

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason Michael; Lessing, Paul Alan; Hoggan, Rita Elaine

In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3Si 2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3Si 2 has been optimized and high phase purity U 3Si 2 has been successfully produced. Results are presentedmore » from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3. Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.« less

Briefing Book. Volume 1: The Evolution of the Nuclear Non-Proliferation Regime (Fourth Edition).

DTIC Science & Technology

1998-01-01

usually termed) nuclear reactors. The first of these is that they contain a core or mass of fissile material (the fuel ) which may weigh tens of tons... HTGR is cooled with helium gas and moderated with graphite. Highly enriched uranium is used as fuel (93 per cent U-235), though this may be mixed with...to convert U-238 in a blanket around the core into Pu-239 at a rate faster than its own consumption of fissile material. They thus produce more fuel

An Interactive Code for a Pressurized Water Reactor Incorporating Temperature and Xenon Feedback.

DTIC Science & Technology

1980-06-01

z2~ 0 z Zzc 0 z zc 0 V) W zAW 22 9- 2 9- 49 8. SMULTI WFFR P𔄂RRA RCWTERRERCT OR - REACTIVITT RESPONSE LGN a RHif a £owt uflif CL TEMPERRTURE...Y .W07l-LltUjW-"UD L-A .4 W 1- *S. l- li -11 0f .. it-4 000 011 4 .WC04LJ W L’ 0 au . U tUN ... ’U * ’.O - J4I 0 S -. I-~- -.44JIUOI I- .,.41..TX...J0

Nuclear Analysis

NASA Technical Reports Server (NTRS)

Clement, J. D.; Kirby, K. D.

1973-01-01

Exploratory calculations were performed for several gas core breeder reactor configurations. The computational method involved the use of the MACH-1 one dimensional diffusion theory code and the THERMOS integral transport theory code for thermal cross sections. Computations were performed to analyze thermal breeder concepts and nonbreeder concepts. Analysis of breeders was restricted to the (U-233)-Th breeding cycle, and computations were performed to examine a range of parameters. These parameters include U-233 to hydrogen atom ratio in the gaseous cavity, carbon to thorium atom ratio in the breeding blanket, cavity size, and blanket size.

Computer program FPIP-REV calculates fission product inventory for U-235 fission

NASA Technical Reports Server (NTRS)

Brown, W. S.; Call, D. W.

1967-01-01

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

76 FR 31379 - Notice of Issuance of Bulletin 2011-01, Mitigating Strategies

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-31

..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. Telephone: 301-415-2963; e-mail... NUCLEAR REGULATORY COMMISSION [NRC-2011-0120] Notice of Issuance of Bulletin 2011-01, Mitigating Strategies AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Notice of Issuance. SUMMARY: The U.S. Nuclear...

Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

NASA Astrophysics Data System (ADS)

Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

2018-02-01

TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

75 FR 68629 - Massachusetts Institute of Technology Reactor Notice of Issuance of Renewed Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-08

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-020; NRC-2010-0313] Massachusetts Institute of Technology Reactor Notice of Issuance of Renewed Facility Operating; License No. R-37 The U.S. Nuclear... Institute of Technology (the licensee), which authorizes continued operation of the Massachusetts Institute...

NRC Targets University Reactors.

ERIC Educational Resources Information Center

Marshall, Eliot

1984-01-01

The Nuclear Regulatory Commission (NRC) wants universities to convert to low-grade fuel in their research reactions. Researchers claim the conversion, which will bring U.S. reactors in line with a policy the NRC is trying to impress on foreigners, could be financially and scientifically costly. Impact of the policy is considered. (JN)

78 FR 9745 - Kewaunee Power Station; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-11

... FURTHER INFORMATION CONTACT: Karl Feintuch, Project Manager, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2013-03037 Filed 2-8-13; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket No. 50-305; NRC-2013-0028] Kewaunee Power Station...

Determination of neutron multiplication coefficients for fuel elements irradiated by spallation neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bhatia, Chitra; Kumar, V.

2010-02-15

A neutron multiplication coefficient, k{sub eff}, has been estimated for spallation neutron flux using the data of spectrum average cross sections of all absorption, fission, and nonelastic reaction channels of {sup 232}Th, {sup 238}U, {sup 235}U, and {sup 233}U fuel elements. It has been revealed that in spallation neutron flux (i) nonfission, nonabsorption reactions play an important role in the calculation of k{sub eff}, (ii) one can obtain a high value of k{sub eff} even for fertile {sup 232}Th fuel, which is hardly possible in a conventional fast reactor, and (iii) spectrum average absorption cross sections of neutron poisons ofmore » a conventional reactor are relatively very small.« less

A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

NASA Astrophysics Data System (ADS)

Bushuev, A. V.; Kozhin, A. F.; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E.

2016-12-01

An active neutron method for measuring the residual mass of 235U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual 235U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of 238U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

Innovative Clean Coal Technology (ICCT). Demonstration of Selective Catalytic Reduction (SCR) technology for the control of nitrogen oxide (NO{sub x}) emissions from high-sulfur coal-fired boilers: Volume 1. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-10-01

The objective of this project is to demonstrate and evaluate commercially available Selective Catalytic Reduction (SCR) catalysts from U.S., Japanese and European catalyst suppliers on a high-sulfur U.S. coal-fired boiler. SCR is a post-combustion nitrogen oxide (NO.) control technology that involves injecting ammonia into the flue gas generated from coal combustion in an electric utility boiler. The flue gas containing ammonia is then passed through a reactor that contains a specialized catalyst. In the presence of the catalyst, the ammonia reacts with NO. to convert it to nitrogen and water vapor. Although SCR is widely practiced in Japan and Europemore » on gas-, oil-, and low-sulfur coal- fired boilers, there are several technical uncertainties associated with applying SCR to U.S. coals. These uncertainties include: 1) potential catalyst deactivation due to poisoning by trace metal species present in U.S. coals that are not present in other fuels. 2) performance of the technology and effects on the balance-of- plant equipment in the presence of high amounts of SO{sub 2} and SO{sub 3}. 3) performance of a wide variety of SCR catalyst compositions, geometries and methods of manufacturer under typical high-sulfur coal-fired utility operating conditions. These uncertainties were explored by operating nine small-scale SCR reactors and simultaneously exposing different SCR catalysts to flue gas derived from the combustion of high sulfur U.S. coal. In addition, the test facility operating experience provided a basis for an economic study investigating the implementation of SCR technology.« less

NRC Licensing Status Summary Report for NGNP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moe, Wayne Leland; Kinsey, James Carl

2014-11-01

The Next Generation Nuclear Plant (NGNP) Project, initiated at Idaho National Laboratory (INL) by the U.S. Department of Energy (DOE) pursuant to provisions of the Energy Policy Act of 2005, is based on research and development activities supported by the Department of Energy Generation IV Nuclear Energy Systems Initiative. The principal objective of the NGNP Project is to support commercialization of high temperature gas-cooled reactor (HTGR) technology. The HTGR is a helium-cooled and graphite moderated reactor that can operate at temperatures much higher than those of conventional light water reactor (LWR) technologies. The NGNP will be licensed for construction andmore » operation by the Nuclear Regulatory Commission (NRC). However, not all elements of current regulations (and their related implementation guidance) can be applied to HTGR technology at this time. Certain policies established during past LWR licensing actions must be realigned to properly accommodate advanced HTGR technology. A strategy for licensing HTGR technology was developed and executed through the cooperative effort of DOE and the NRC through the NGNP Project. The purpose of this report is to provide a snapshot of the current status of the still evolving pre-license application regulatory framework relative to commercial HTGR technology deployment in the U.S. The following discussion focuses on (1) describing what has been accomplished by the NGNP Project up to the time of this report, and (2) providing observations and recommendations concerning actions that remain to be accomplished to enable the safe and timely licensing of a commercial HTGR facility in the U.S.« less

Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loflin, Leonard; McRimmon, Beth

2014-12-18

This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

Nondestructive assay of EBR-II blanket elements using resonance transmission analysis

NASA Astrophysics Data System (ADS)

Klann, Raymond Todd

1998-10-01

Resonance transmission analysis utilizing a filtered reactor beam was examined as a means of determining the 239Pu content in Experimental Breeder Reactor - II depleted uranium blanket elements. The technique uses cadmium and gadolinium filters along with a 239Pu fission chamber to isolate the 0.3 eV resonance in 239Pu. In the energy range of this resonance (0.1 eV to 0.5 eV), the total microscopic cross-section of 239Pu is significantly greater than the cross- sections of 238U and 235U. This large difference allows small changes in the 239Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and 239Pu foils indicate a significant change in response based on the 239Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of 239Pu up to approximately two weight percent.

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less

76 FR 81992 - PPL Bell Bend, LLC; Combined License Application for Bell Bend Nuclear Power Plant; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-29

... License Application for Bell Bend Nuclear Power Plant; Exemption 1.0 Background PPL Bell Bend, LLC... for Nuclear Power Plants.'' This reactor is to be identified as Bell Bend Nuclear Power Plant (BBNPP... based upon the U.S. EPR reference COL (RCOL) application for UniStar's Calvert Cliffs Nuclear Power...

Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1976

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1977-08-02

This bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1976. The report includes 1,253 abstracts that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes ofmore » failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Three of the unique events that occurred during the year are reviewed in detail.« less

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

Fabrication of (U,Am)O2 pellet with controlled porosity from oxide microspheres

NASA Astrophysics Data System (ADS)

Ramond, Laure; Coste, Philippe; Picart, Sébastien; Gauthé, Aurélie; Bataillea, Marc

2017-08-01

U1-xAmxO2±δ mixed-oxides are considered as promising compounds for americium heterogeneous transmutation in Sodium Fast Neutron Reactor. Porous microstructure is envisaged in order to facilitate helium and fission gas release and to reduce pellet swelling during irradiation and under self-irradiation. In this study, the porosity is created by reducing (U,Am)3O8 microspheres into (U,Am)O2 during the sintering. This reduction is accompanied by a decrease of the lattice volume that leads to the creation of open porosity. Finally, an (U0.90Am0.10)O2 porous ceramic pellet (D∼89% of the theoretical density TD) with controlled porosity (≥8% open porosity) was obtained from mixed-oxide microspheres obtained by the Weak Acid Resin (WAR) process.

130. ARAII Administration building (ARA613) vicinity map and plot plan ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

130. ARA-II Administration building (ARA-613) vicinity map and plot plan showing relationship to other existing buildings on site and to ARA-602, to which this building was attached. F.C. Torkelson Comapny 842-area/SL-1-101-U-1. Date: October 1959. Ineel index code no. 070-0101-65-851-150053. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

NASA Astrophysics Data System (ADS)

Metzger, Kathryn E.

Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it difficult to predict fuel-cladding mechanical behavior. This information is essential for designing accident tolerant fuel systems where ceramic claddings, like silicon carbide (SiC) are proposed. This research provides a model for both the thermal and irradiation creep behavior for U3Si2. This body of research is comprised of both experimental and modeling components. Characterization of the fuel microstructure includes; optical microscopy with pore and grain size analysis, helium pycnometry for density determination, mercury intrusion porosimetry, compositional analysis in the form of XRD, second phase identification using EDX, electrical resistance measurement via four point probe, determination of hardness and toughness through Vickers indentation testing, and determination of elastic properties using the impulse excitation method. Post-sintering grain size data allowed for the determination of grain boundary activation energy and diffusion coefficients, which were used to develop creep models. This was extended to lattice and irradiation enhanced diffusion in order to develop a U3Si2 creep model over thermal and irradiation creep regimes. In addition to the creep model, thermal and swelling behavior models for U3Si2 were implemented into the BISON fuel performance code. A series of simulations evaluated the performance and behavior of U3Si2 under typical light water reactor conditions with advanced SiC ceramic cladding. Simulation results show that fuel creep relieves stress in the ceramic cladding and postpones the. moment of fuel-clad contact. However, the stress reduction to the cladding is minimal because the fuel creep rate is low while the swelling rate is high. Future work should include the investigation of monolithic U3Si2 irradiation swelling since the current model relies upon the swelling data of U3Si2 particles in a metallic dispersion fuel. Additionally, planned thermal creep testing at the University of South Carolina can provide confirmation of the U3Si2 creep model contained herein.

TEMPERATURE COEFFICIENTS OF HETEROGENEOUS U-238-U-235 FUELED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Astley, E.R.; Mansius, C.A.

1958-05-14

An analytical method of determining the effective reactivity coefficient from fundamental cross sections using the four factor formula is presented. Values of the coefficient obtained by this method compare well with experiment. (A.C.)

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, M. L.; Peko, D.; Farmer, M.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, M. L.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

NASA Astrophysics Data System (ADS)

Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

2017-07-01

FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

Irradiation testing of high density uranium alloy dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups ofmore » 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.« less

In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haghighi, M. H.; Kring, C. T.; McGehee, J. T.

2002-02-26

The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 tomore » December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using the uranium deposit removal system and equipment. After removal a series of NDA measurements was performed to determine the amount of uranium material remaining in the ACB, the amount of uranium material removed from the ACB, and the amount of uranium material remaining in the uranium removal equipment due to removal activities.« less

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Theoretical Interpretation of the Measurement of Diffusion Parameters with Pulsed Neutron Source; INTERPRETAZIONE TEORICA DELLE MISURE DI PARAMETRI DI DIFFUSIONE COL METODO DELLE SORGENTI NEUTRONICHE PULSATE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boffi, V.C.; Molinari, V.G.; Parks, D.E.

1962-05-01

Features of the pulsed neution source theory connected with the measurement of diffusion parameters are discussed. Various analytical procedures for determining the decay constant of the fully thermalized neutron flux are compared. The problem of the diffusion coefficient definition is also considered in some detail. (auth)

New tools for subsurface imaging of 3D seismic Node data in hydrocarbon exploration =

NASA Astrophysics Data System (ADS)

Benazzouz, Omar

A aquisicao de dados sismicos de reflexao multicanal 3D/4D usando Ocean Bottom NODES de 4 componentes constitui atualmente um sector de importancia crescente no mercado da aquisicao de dados reflexao sismica marinha na industria petrolifera. Este tipo de dados permite obter imagens de sub-superficie de alta qualidade, com baixos niveis de ruido, banda larga, boa iluminacao azimutal, offsets longos, elevada resolucao e aquisicao de tanto ondas P como S. A aquisicao de dados e altamente repetitiva e portanto ideal para campanhas 4D. No entanto, existem diferencas significativas na geometria de aquisicao e amostragem do campo de ondas relativamente aos metodos convencionais com streamers rebocados a superficie, pelo que e necessario desenvolver de novas ferramentas para o processamento deste tipo de dados. Esta tese investiga tres aspectos do processamento de dados de OBSs/NODES ainda nao totalmente resolvidos de forma satisfatoria: a deriva aleatoria dos relogios internos, o posicionamento de precisao dos OBSs e a implementacao de algoritmos de migracao prestack 3D em profundidade eficientes para obtencao de imagens precisas de subsuperficie. Foram desenvolvidos novos procedimentos para resolver estas situacoes, que foram aplicados a dados sinteticos e a dados reais. Foi desenvolvido um novo metodo para deteccao e correccao de deriva aleatoria dos relogios internos, usando derivadas de ordem elevada. Foi ainda desenvolvido um novo metodo de posicionamento de precisao de OBSs usando multilateracao e foram criadas ferramentas de interpolacao/extrapolacao dos modelos de velocidades 3D de forma a cobrirem a extensao total area de aquisicao. Foram implementados algoritmos robustos de filtragem para preparar o campo de velocidades para o tracado de raios e minimizar os artefactos na migracao Krichhoff pre-stack 3D em profundidade. Os resultados obtidos mostram um melhoramento significativo em todas as situacoes analisadas. Foi desenvolvido o software necessario para o efeito e criadas solucoes computacionais eficientes. As solucoes computacionais desenvolvidas foram integradas num software standard de processamento de sismica (SPW) utilizado na industria, de forma a criar, conjuntamente com as ferramentas ja existentes, um workflow de processamento integrado para dados de OBS/NODES, desde a aquisicao e controle de qualidade a producao dos volumes sismicos migrados pre-stack em profundidade.

Shield materials recommended for space power nuclear reactors

NASA Technical Reports Server (NTRS)

Kaszubinski, L. J.

1973-01-01

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Processing of U-2.5Zr-7.5Nb and U-3Zr-9Nb alloys by sintering process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dos Santos, A. M. M.; Ferraz, W. B.; Lameiras, F. S.

2012-07-01

To minimize the risk of nuclear proliferation, there is worldwide interest in reducing fuel enrichment of research and test reactors. To achieve this objective while still guaranteeing criticality and cycle length requirements, there is need of developing high density uranium metallic fuels. Alloying elements such as Zr, Nb and Mo are added to uranium to improve fuel performance in reactors. In this context, the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) is developing the U-2.5Zr-7.5Nb and U-3Zr-9Nb (weight %) alloys by the innovative process of sintering that utilizes raw materials in the form of powders. The powders were pressed atmore » 400 MPa and then sintered under a vacuum of about 1x10{sup -4} Torr at temperatures ranging from 1050 deg. to 1500 deg.C. The densities of the alloys were measured geometrically and by hydrostatic method and the phases identified by X ray diffraction (XRD). The microstructures of the pellets were observed by scanning electron microscopy (SEM) and the alloying elements were analyzed by energy dispersive X-ray spectroscopy (EDS). The results obtained showed the fuel density to slightly increase with the sintering temperature. The highest density achieved was approximately 80% of theoretical density. It was observed in the pellets a superficial oxide layer formed during the sintering process. (authors)« less

Characterizing the biotransformation of sulfur-containing wastes in simulated landfill reactors.

PubMed

Sun, Wenjie; Sun, Mei; Barlaz, Morton A

2016-07-01

Landfills that accept municipal solid waste (MSW) in the U.S. may also accept a number of sulfur-containing wastes including residues from coal or MSW combustion, and construction and demolition (C&D) waste. Under anaerobic conditions that dominate landfills, microbially mediated processes can convert sulfate to hydrogen sulfide (H2S). The presence of H2S in landfill gas is problematic for several reasons including its low odor threshold, human toxicity, and corrosive nature. The objective of this study was to develop and demonstrate a laboratory-scale reactor method to measure the H2S production potential of a range of sulfur-containing wastes. The H2S production potential was measured in 8-L reactors that were filled with a mixture of the target waste, newsprint as a source of organic carbon required for microbial sulfate reduction, and leachate from decomposed residential MSW as an inoculum. Reactors were operated with and without N2 sparging through the reactors, which was designed to reduce H2S accumulation and toxicity. Both H2S and CH4 yields were consistently higher in reactors that were sparged with N2 although the magnitude of the effect varied. The laboratory-measured first order decay rate constants for H2S and CH4 production were used to estimate constants that were applicable in landfills. The estimated constants ranged from 0.11yr(-1) for C&D fines to 0.38yr(-1) for a mixed fly ash and bottom ash from MSW combustion. Copyright © 2016 Elsevier Ltd. All rights reserved.

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

NUCLEAR REACTOR

DOEpatents

Starr, C.

1963-01-01

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)

76 FR 26775 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on EPR; Cancellation to...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-09

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on EPR; Cancellation to May 11, 2011, ACRS Meeting-- Federal Register Notice The Federal Register Notice for the ACRS Subcommittee Meeting on the design certification application review of the U.S...

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2009-11-05

exports to India in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor Export Deal,” Chennai...agreements. In 1974, P.L. 93-485 amended Section 123 d. to include agreements that covered reactors producing more than 5 MW thermal or special nuclear

76 FR 23339 - Notice of Issuance of License Amendment Regarding Decommission Plan Approval; University of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-26

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-113; NRC-2009-0549] Notice of Issuance of License Amendment Regarding Decommission Plan Approval; University of Arizona Research Reactor The U.S. Nuclear... located within the University of Arizona Nuclear Reactor Laboratory (NRL) on the 325-acre campus of the...

76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

Patents - Harold Clayton Urey

Science.gov Websites

) US 2,947,472 CENTRIFUGE APPARATUS - Urey, H. C.; Skarstrom, C; Cohen, K; August 2, 1960 (to U. S Commission) This patent is concerned with a heavy water enriched uranium power reactor capable of producing reactor where the stream from both reaction zone and absorber zone is separated from the liquid and solid

75 FR 16517 - FirstEnergy Nuclear Operating Company; Environmental Assessment and Finding of No Significant Impact

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

... Perry. FOR FURTHER INFORMATION CONTACT: Michael Mahoney, Office of Nuclear Reactor Regulation, U.S... Nuclear Reactor Regulation. [FR Doc. 2010-7331 Filed 3-31-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-440; NRC-2010-0124] FirstEnergy Nuclear Operating...

PLUTONIUM-THORIUM ALLOYS

DOEpatents

Schonfeld, F.W.

1959-09-15

New plutonium-base binary alloys useful as liquid reactor fuel are described. The alloys consist of 50 to 98 at.% thorium with the remainder plutonium. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are easy fabrication, phase stability, and the accompanying advantuge of providing a means for converting Th/sup 232/ into U/sup 233/.

The WSTIAC Quarterly. Volume 9, Number 3

DTIC Science & Technology

2010-01-25

program .[8] THE THORIUM FUEL CYCLE AND LFTR POWER PLANT The thorium fuel cycle is based on a series of neutron absorp- tion and beta decay processes...the fig- ure is a graphite matrix moderated MSR reactor with fuel salt mixture (ThF4-U233F4) being circulated by a pump through the core and to a...the core as purified salt. As one of the unique safety features, a melt-plug at the reactor bottom would permit the reactor fluid fuel to be drained

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

System Study: Reactor Core Isolation Cooling 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

NUCLEAR REACTOR

DOEpatents

Young, G.

1963-01-01

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

Development and Characterization of 6Li-doped Liquid Scintillator Detectors for PROSPECT

NASA Astrophysics Data System (ADS)

Gaison, Jeremy; Prospect Collaboration

2016-09-01

PROSPECT, the Precision Reactor Oscillation and Spectrum experiment, is a phased reactor antineutrino experiment designed to search for eV-scale sterile neutrinos via short-baseline neutrino oscillations and to make a precision measurement of the 235U reactor antineutrino spectrum. A multi-ton, optically segmented detector will be deployed at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR) to measure the reactor spectrum at baselines ranging from 7-12m. A two-segment detector prototype with 50 liters of active liquid scintillator target has been built to verify the detector design and to benchmark its performance. In this presentation, we will summarize the performance of this detector prototype and describe the optical and energy calibration of the segmented PROSPECT detectors.

First-principles study of the surface properties of U-Mo system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

U-Mo alloys are promising fuels for future high-performance research reactors with low enriched uranium. Surface properties, such as surface energy, are important inputs for mesoscale simulations (e.g., phase field method) of fission gas bubble behaviors in irradiated nuclear fuels. The lack of surface energies of U-Mo alloys prevents an accurate modeling of the morphology of gas bubbles and gas bubble-induced fuel swelling. To this end, we study the surface properties of U-Mo system, including bcc Mo, alpha-U, gamma-U, and gamma U-Mo alloys. All surfaces up to a maximum Miller index of three and two are calculated for cubic Mo andmore » gamma-U and non-cubic alpha-U, respectively. The equilibrium crystal shapes of bcc Mo, alpha-U and gamma-U are constructed using the calculated surface energies. The dominant surface orientations and the area fraction of each facet are determined from the constructed equilibrium crystal shape. The disordered gamma U-Mo alloys are simulated using the Special Quasirandom Structure method. The (1 1 0) and (1 0 0) surface energies of gamma U-7Mo and U-10Mo alloys are predicted to lie between those of gamma-U and bcc Mo, following a linear combination of the two constituents' surface energies. To better compare with future measurements of surface energies, the area fraction weighted surface energies of alpha-U, gamma-U and gamma U-7Mo and U-10Mo alloys are also predicted. (C) 2017 Published by Elsevier B.V.« less

Radio-toxicity of spent fuel of the advanced heavy water reactor.

PubMed

Anand, S; Singh, K D S; Sharma, V K

2010-01-01

The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

The role of nuclear reactors in space exploration and development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built and flew space reactors; it is time to do so again.« less

Speciation and reactivity of uranium products formed during in situ bioremediation in a shallow alluvial aquifer.

PubMed

Alessi, Daniel S; Lezama-Pacheco, Juan S; Janot, Noémie; Suvorova, Elena I; Cerrato, José M; Giammar, Daniel E; Davis, James A; Fox, Patricia M; Williams, Kenneth H; Long, Philip E; Handley, Kim M; Bernier-Latmani, Rizlan; Bargar, John R

2014-11-04

In this study, we report the results of in situ U(VI) bioreduction experiments at the Integrated Field Research Challenge site in Rifle, Colorado, USA. Columns filled with sediments were deployed into a groundwater well at the site and, after a period of conditioning with groundwater, were amended with a mixture of groundwater, soluble U(VI), and acetate to stimulate the growth of indigenous microorganisms. Individual reactors were collected as various redox regimes in the column sediments were achieved: (i) during iron reduction, (ii) just after the onset of sulfate reduction, and (iii) later into sulfate reduction. The speciation of U retained in the sediments was studied using X-ray absorption spectroscopy, electron microscopy, and chemical extractions. Circa 90% of the total uranium was reduced to U(IV) in each reactor. Noncrystalline U(IV) comprised about two-thirds of the U(IV) pool, across large changes in microbial community structure, redox regime, total uranium accumulation, and reaction time. A significant body of recent research has demonstrated that noncrystalline U(IV) species are more suceptible to remobilization and reoxidation than crystalline U(IV) phases such as uraninite. Our results highlight the importance of considering noncrystalline U(IV) formation across a wide range of aquifer parameters when designing in situ remediation plans.

Speciation and Reactivity of Uranium Products Formed during in Situ Bioremediation in a Shallow Alluvial Aquifer

PubMed Central

2015-01-01

In this study, we report the results of in situ U(VI) bioreduction experiments at the Integrated Field Research Challenge site in Rifle, Colorado, USA. Columns filled with sediments were deployed into a groundwater well at the site and, after a period of conditioning with groundwater, were amended with a mixture of groundwater, soluble U(VI), and acetate to stimulate the growth of indigenous microorganisms. Individual reactors were collected as various redox regimes in the column sediments were achieved: (i) during iron reduction, (ii) just after the onset of sulfate reduction, and (iii) later into sulfate reduction. The speciation of U retained in the sediments was studied using X-ray absorption spectroscopy, electron microscopy, and chemical extractions. Circa 90% of the total uranium was reduced to U(IV) in each reactor. Noncrystalline U(IV) comprised about two-thirds of the U(IV) pool, across large changes in microbial community structure, redox regime, total uranium accumulation, and reaction time. A significant body of recent research has demonstrated that noncrystalline U(IV) species are more suceptible to remobilization and reoxidation than crystalline U(IV) phases such as uraninite. Our results highlight the importance of considering noncrystalline U(IV) formation across a wide range of aquifer parameters when designing in situ remediation plans. PMID:25265543

Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D., Jr.; Miller, B. D.; Wachs, D. M.; Allen, T. R.; Kirk, M.; Rest, J.

2011-04-01

The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al) 3, (U, Mo)(Si, Al) 3, UMo 2Al 20, U 6Mo 4Al 43, and UAl 4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to ˜100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al) 3 and UAl 4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al) 3 and UMo 2Al 20 phases become amorphous at 1 and ˜2 dpa, respectively, and show no evidence of voids at 100 dpa. The U 6Mo 4Al 43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

New reactor technology: safety improvements in nuclear power systems.

PubMed

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Probabilistic pipe fracture evaluations for leak-rate-detection applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

The MSFR as a flexible CR reactor: the viewpoint of safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fiorina, C.; Cammi, A.; Franceschini, F.

2013-07-01

In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal tomore » about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. βeff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of βeff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.« less

Continuous production of ethanol from starch using glucoamylase and yeast co-immobilized in pectin gel.

PubMed

Giordano, Raquel L C; Trovati, Joubert; Schmidell, Willibaldo

2008-03-01

This work presents a continuous simultaneous saccharification and fermentation (SSF) process to produce ethanol from starch using glucoamylase and Saccharomyces cerevisiae co-immobilized in pectin gel. The enzyme was immobilized on macroporous silica, after silanization and activation of the support with glutaraldehyde. The silica-enzyme derivative was co-immobilized with yeast in pectin gel. This biocatalyst was used to produce ethanol from liquefied manioc root flour syrup, in three fixed bed reactors. The initial reactor yeast load was 0.05 g wet yeast/ml of reactor (0.1 g wet yeast/g gel), used in all SSF experiments. The enzyme concentration in the reactor was defined by running SSF batch assays, using different amount of silica-enzyme derivative, co-immobilized with yeast in pectin gel. The chosen reactor enzyme concentration, 3.77 U/ml, allowed fermentation to be the rate-limiting step in the batch experiment. In this condition, using initial substrate concentration of 166.0 g/l of total reducing sugars (TRS), 1 ml gel/1 ml of medium, ethanol productivity of 8.3 g/l/h was achieved, for total conversion of starch to ethanol and 91% of the theoretical yield. In the continuous runs, feeding 163.0 g/l of TRS and using the same enzyme and yeast concentrations used in the batch run, ethanol productivity was 5.9 g ethanol/l/h, with 97% of substrate conversion and 81% of the ethanol theoretical yield. Diffusion effects in the extra-biocatalyst film seemed to be reduced when operating at superficial velocities above 3.7 x 10(-4) cm/s.

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sears, D.F.; Primeau, M.F.; Buchanan, C.

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinationsmore » showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.« less

U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prabhakaran, Ramprashad

The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, highmore » uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.« less

Fukushima Daiichi Information Repository FY13 Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Curtis; Phelan, Cherie; Schwieder, Dave

The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in commercial nuclear power plant operating history. Much will be learned that may be applicable to the U.S. reactor fleet, nuclear fuel cycle facilities, and supporting systems, and the international reactor fleet. For example, lessons from Fukushima Daiichi may be applied to emergency response planning, reactor operator training, accident scenario modeling, human factors engineering, radiation protection, and accident mitigation; as well as influence U.S. policies towards the nuclear fuel cycle including power generation, and spent fuel storage, reprocessing, and disposal. This document describesmore » the database used to establish a centralized information repository to store and manage the Fukushima data that has been gathered. The data is stored in a secured (password protected and encrypted) repository that is searchable and available to researchers at diverse locations.« less

Soviet space nuclear reactor incidents - Perception versus reality

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

A thermal conductivity model for U-­Si compounds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Yongfeng; Andersson, Anders David Ragnar

U 3Si 2 is a candidate for accident tolerant nuclear fuel being developed as an alternative to UO 2 in commercial light water reactors (LWRs). One of its main benefits compared to UO 2 is higher thermal conductivity that increases with temperature. This increase is contrary to UO 2, for which the thermal conductivity decreases with temperature. The reason for the difference is the electronic origin of thermal conductivity in U 3Si 2, as compared to the phonon mechanism responsible for thermal transport in UO 2. The phonon thermal conductivity in UO 2 is unusually low for a fluorite oxidemore » due to the strong interaction with the spins in the paramagnetic phase. The thermal conductivity of U 3Si 2 as well as other U-­Si compounds has been measured experimentally [1-­4]. However, for fuel performance simulations it is also critical to model the degradation of the thermal conductivity due to damage and microstructure evolution caused by the reactor environment (irradiation and high temperature). For UO 2 this reduction is substantial and it has been the topic of extensive NEAMS research resulting in several publications [5, 6]. There are no data or models for the evolution of the U 3Si 2 thermal conductivity under irradiation. We know that the intrinsic thermal conductivities of UO 2 (semi-conductor) and U 3Si 2 (metal) are very different, and we do not necessarily expect the dependence on damage to be the same either, which could present another advantage for the silicide fuel. In this report we summarize the first step in developing a model for the thermal conductivity of U-­Si compounds with the goal of capturing the effect of damage in U 3Si 2. Next year, we will focus on lattice damage. We will also attempt to assess the impact of fission gas bubbles.« less

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2012 CFR

2012-01-01

...) of 10 CFR 50.34—Post-Accident Sampling for Boron, Chloride, and Dissolved Gases; and 3. Paragraph (f... affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if: (1) There is a substantial increase in the probability of an ex-vessel...

Boron neutron capture therapy induces apoptosis of glioma cells through Bcl-2/Bax

PubMed Central

2010-01-01

Background Boron neutron capture therapy (BNCT) is an alternative treatment modality for patients with glioma. The aim of this study was to determine whether induction of apoptosis contributes to the main therapeutic efficacy of BNCT and to compare the relative biological effect (RBE) of BNCT, γ-ray and reactor neutron irradiation. Methods The neutron beam was obtained from the Xi'an Pulsed Reactor (XAPR) and γ-rays were obtained from [60Co] γ source of the Fourth Military Medical University (FMMU) in China. Human glioma cells (the U87, U251, and SHG44 cell lines) were irradiated by neutron beams at the XAPR or [60Co] γ-rays at the FMMU with different protocols: Group A included control nonirradiated cells; Group B included cells treated with 4 Gy of [60Co] γ-rays; Group C included cells treated with 8 Gy of [60Co] γ-rays; Group D included cells treated with 4 Gy BPA (p-borono-phenylalanine)-BNCT; Group E included cells treated with 8 Gy BPA-BNCT; Group F included cells irradiated in the reactor for the same treatment period as used for Group D; Group G included cells irradiated in the reactor for the same treatment period as used for Group E; Group H included cells irradiated with 4 Gy in the reactor; and Group I included cells irradiated with 8 Gy in the reactor. Cell survival was determined using the 3-(4,5-dimethylthiazol-2-yl-2,5-diphenyltetrazolium (MTT) cytotoxicity assay. The morphology of cells was detected by Hoechst33342 staining and transmission electron microscope (TEM). The apoptosis rate was detected by flow cytometer (FCM). The level of Bcl-2 and Bax protein was measured by western blot analysis. Results Proliferation of U87, U251, and SHG44 cells was much more strongly inhibited by BPA-BNCT than by irradiation with [60Co] γ-rays (P < 0.01). Nuclear condensation was determined using both a fluorescence technique and electron microscopy in all cell lines treated with BPA-BNCT. Furthermore, the cellular apoptotic rates in Group D and Group E treated with BPA-BNCT were significantly higher than those in Group B and Group C irradiated by [60Co] γ-rays (P < 0.01). The clonogenicity of glioma cells was reduced by BPA-BNCT compared with cells treated in the reactor (Group F, G, H, I), and with the control cells (P < 0.01). Upon BPA-BNCT treatment, the Bax level increased in glioma cells, whereas Bcl-2 expression decreased. Conclusions Compared with γ-ray and reactor neutron irradiation, a higher RBE can be achieved upon treatment of glioma cells with BNCT. Glioma cell apoptosis induced by BNCT may be related to activation of Bax and downregulation of Bcl-2. PMID:21122152

Modification of apparent fission yields by Chemical Fractionation following Fission (CFF)

NASA Astrophysics Data System (ADS)

Hohenberg, Charles; Meshik, Alex

2008-04-01

Grain-by-grain studies of the 2 billion year old Oklo natural reactor, using laser micro-extraction^1,2, yield detailed information about Oklo, a water-moderated pulsed reactor, cycle times, total neutron fluence and duration, but it also demonstrates Chemical Fractionation following Fission. In the CFF process, members of an isobaric yield chain with long half-lives are subject to migration before decay can occur. Of particular interest is the 129 isobar where 17 million ^129I can migrate out of the host grain before decay, and iodine compounds are water soluble. This is amply demonstated by the variation of Xe spectra between micron-sized uranium-bearing minerals and adjacent uranium-free minerals. Fission 129 yields for the spontaneous fission of ^238U generally come from measured ^129Xe in pitchblend^2, ores emplaced by aqueous activity, and are incorrect due to the CFF process. ^238U yields for the 131 and 129 chains, reported in Hyde^3, as 0.455 +- .02 and < 0.012, respectively, the latter being anomalously low. ^1A Meshik, C Hohenberg and O Pravdivtesva, PRL 93, 182302 (2004); A Meshik Sci. Am. Nov (2005), 55; ^2E K Hyde, Nucl Prop of Heavy Elements III (1964).

Expeditious method to determine uranium in the process control samples of chemical plant separating (233)U from thoria irradiated in power reactors.

PubMed

Kedari, C S; Kharwandikar, B K; Banerjee, K

2016-11-01

Analysis of U in the samples containing a significant proportion of (232)U and high concentration of Th is of great concern. Transmutation of Th in the nuclear power reactor produces a notable quantity of (232)U (half life 68.9 years) along with fissile isotope (233)U. The decay series of (232)U is initiated with (228)Th (half life 1.9 year) and it is followed by several short lived α emitting progenies, (224)Ra, (220)Rn, (216)Po, (212)Bi and (212)Po. Even at the smallest contamination of (228)Th in the sample, a very high pulse rate of α emission is obtained, which is to be counted for the radiometric determination of [U]. A commercially available anionic type of extractant Alamine®336 is used to obtain the selective extraction of U from other alpha active elements and fission products present in the sample. Experimental conditions of liquid-liquid extraction (LLE) are optimized for obtaining maximum decontamination and recovery of U in the organic phase. The effect of some interfering ionic impurities in the sample on the process of separation is investigated. Depending on the level of the concentration of U in the samples, spectrophotometry or radiometry methods are adopted for its determination after separation by LLE. Under optimized experimental conditions, i.e. 5.5M HCl in the aqueous phase and 0.27M Alamin®336 in the organic phase, the recovery of U is about 100%, the decontamination factor with respect to Th is >2000 and the extraction of fission products like (90)Sr, (144)Ce and (134,137)Cs is negligible. The detection limit for [U] using α radiometry is 10mg/L, even in presence of >100g/L of Th in the sample. Accuracy and precision for the determination of U is also assessed. Reproducibility of results is within 5%. This method shows very good agreement with the results obtained by mass spectrometry. Copyright © 2016 Elsevier B.V. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mitchell K Meyer

Blister–threshold testing of fuel plates is a standard method through which the safety margin for operation of plate-type in research and test reactors is assessed. The blister-threshold temperature is indicative of the ability of fuel to operate at high temperatures for short periods of time (transient conditions) without failure. This method of testing was applied to the newly developed U-Mo monolithic fuel system. Blister annealing studies on the U-Mo monolithic fuel plates began in 2007, with the Reduced Enrichment for Research and Test Reactors (RERTR)-6 experiment, and they have continued as the U-Mo fuel system has evolved through the researchmore » and development process. Blister anneal threshold temperatures from early irradiation experiments (RERTR-6 through RERTR-10) ranged from 400 to 500°C. These temperatures were projected to be acceptable for NRC-licensed research reactors and the high-power Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) based on current safety-analysis reports (SARs). Initial blister testing results from the RERTR-12 experiment capsules X1 and X2 showed a decrease in the blister-threshold temperatures. Blister threshold temperatures from this experiment ranged from 300 to 400°C. Selected plates from the AFIP-4 experiment, which was fabricated using a process similar to that used to fabricate the RERTR-12 experiment, also underwent blister testing to determine whether results would be similar. The measured blister-threshold temperatures from the AFIP-4 plates fell within the same blister-threshold temperature range measured in the RERTR-12 plates. Investigation of the cause of this decrease in bister threshold temperature is being conducted under the guidance of Idaho National Laboratory PLN-4155, “Analysis of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments,” and is driven by hypotheses. The main focus of the investigation is in the following areas: 1. Fabrication variables 2. Pre-irradiation characterization 3. Irradiation conditions 4. Post-irradiation examination 5. Additional blister testing 6. Mechanical modeling This report documents the preliminary results of this investigation. Several hypotheses can be dismissed as a result of this investigation. Two primary categories of causes remain. The most prominent theory, supported by the data, is that low blister-threshold temperature is the result of mechanical energy imparted on the samples during the fabrication process (hot and cold rolling) without adequate post processing (annealing). The mechanisms are not clearly understood and require further investigation, but can be divided into two categories: • Residual Stress • Undesirable interaction boundary and/or U-Mo microstructure change A secondary theory that cannot be dismissed with the information that is currently available is that a change in the test conditions has resulted in a statistically significant downward shift of measured blister temperature. This report outlines the results of the forensic investigations conducted to date. The data and conclusions presented in this report are preliminary. Definitive cause and effect relationships will be established by future experimental programs.« less

77 FR 4838 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-31

...'' of the U.S. EPR Document Control Design (DCD) Safety Evaluation Report (SER) with Open Items. The...:30 P.M. The Subcommittee will review Chapter 3, ``Design of Structures, Components, Equipment, and...

Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson

2012-06-01

The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fissionmore » rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.« less

Growth and laccase production kinetics of Trametes versicolor in a stirred tank reactor.

PubMed

Thiruchelvam, A T; Ramsay, Juliana A

2007-03-01

White rot fungi are a promising option to treat recalcitrant organic molecules, such as lignin, polycyclic aromatic hydrocarbons, and textile dyes, because of the lignin-modifying enzymes (LMEs) they secrete. Because knowledge of the kinetic parameters is important to better design and operate bioreactors to cultivate these fungi for degradation and/or to produce LME(s), these parameters were determined using Trametes versicolor ATCC 20869 (ATCC, American Type Culture Collection) in a magnetic stir bar reactor. A complete set of kinetic data has not been previously published for this culture. Higher than previously reported growth rates with high laccase production of up to 1,385 U l(-1) occurred during growth without [Formula: see text] or glucose limitation. The maximum specific growth rate averaged 0.94 +/- 0.23 day(-1), whereas the maximum specific substrate consumption rates for glucose and ammonium were 3.37 +/- 1.16 and 0.15 +/- 0.04 day(-1), respectively. The maximum specific oxygen consumption rate was 1.63 +/- 0.36 day(-1).

ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. S. Chang

2007-09-01

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 fuelmore » 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 flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner/outer heat flux more effectively. The optimized LEU relative radial fission heat flux profile is bounded by the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (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, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.« less

Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

NASA Astrophysics Data System (ADS)

Zhou, Hao-Jun; Yin, Yan-Peng; Fan, Xiao-Qiang; Li, Zheng-Hong; Pu, Yi-Kang

2016-06-01

A perturbation method is proposed to obtain the effective delayed neutron fraction β eff of a cylindrical highly enriched uranium reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the reactor reactivity perturbation Δρ of the sample in β eff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation β eff = dk/Δρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average β eff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for β eff can be used to evaluate the reliabilities of the delayed neutron parameters. This work shows that the delayed neutron data of 235U and 238U from G.R. Keepin’s publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2. Supported by Foundation of Key Laboratory of Neutron Physics, China Academy of Engineering Physics (2012AA01, 2014AA01), National Natural Science Foundation (11375158, 91326104)

Prioritized List of Research Needs to support MRWFD Case Study Flowsheet Advancement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Law, Jack Douglas; Soelberg, Nicholas Ray

In FY-13, a case study evaluation was performed of full recycle technologies for both the processing of light-water reactor (LWR) used nuclear fuels as well as fast reactor (FR) fuel in the full recycle option. This effort focused on the identification of the case study processes and the initial preparation of material balance flowsheets for the identified technologies. In identifying the case study flowsheets, it was decided that two cases would be developed: one which identifies the flowsheet as currently developed and another near-term target flowsheet which identifies the flowsheet as envisioned within two years, pending the results of ongoingmore » research. The case study focus is on homogeneous aqueous recycle of the U/TRU resulting from the processing of LWR fuel as feed for metal fuel fabrication. The metal fuel is utilized in a sodium-cooled fast reactor, and the used fast reactor fuel is processed using electrochemical separations. The recovered U/TRU from electrochemical separations is recycled to fuel fabrication and the fast reactor. Waste streams from the aqueous and electrochemical processing are treated and prepared for disposition. Off-gas from the separations and waste processing are also treated. As part of the FY-13 effort, preliminary process unknowns and research needs to advance the near-term target flowsheets were identified. In FY-14, these research needs were updated, expanded and prioritized. This report again updates the prioritized list of research needs based upon results to date in FY-15. The research needs are listed for each of the main portions of the flowsheet: 1) Aqueous headend, 2) Headend tritium pretreatment off-gas, 3) Aqueous U/Pu/Np recovery, 4) Aqueous TRU product solidification, 5) Aqueous actinide/lanthanide separation, 6) Aqueous off-gas treatment, 7) Aqueous HLW management, 8) Treatment of aqueous process wastes, 9) E-chem actinide separations, 10) E-chem off-gas, 11) E-chem HLW management. The identified research needs were prioritized within each of these areas. No effort was made to perform an overall prioritization. This information will be used by the MRWFD Campaign leadership in research planning for FY-16. Additionally, this information will be incorporated into the next version of the Case Study Report scheduled to be issued September 2015.« less

MTR, TRA603. CONTROL ROOM DETAILS. ACOUSTIC PLASTER CEILING, USHAPED CONSOLE, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR, TRA-603. CONTROL ROOM DETAILS. ACOUSTIC PLASTER CEILING, U-SHAPED CONSOLE, INSTRUMENT PANELS, GLASS DOOR, ASPHALT TILE FLOOR AND COLORS. BLAW-KNOX 3150-803-11, 10/1950. INL INDEX NO. 531-0603-00-098-100570, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

General layout of reactor and control areas upon advent of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

General layout of reactor and control areas upon advent of power burst facility (PBF). Shows relationship of PBF to SPERT-I, -II, -III, and -IV. Ebasco Services 1205-PER/PBF-U-102. Date: July 1965. INEEL index no. 761-0100-00-205-123006 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

75 FR 8411 - Office of New Reactors: Interim Staff Guidance on Assessing the Consequences of an Accidental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-24

... the public will be better served by being able to review and comment on both documents at this time... Construction Inspection and Operational Programs, Office of New Reactors, U.S. Nuclear Regulatory Commission... conditions for such releases and define acceptable assumptions to describe exposure scenarios and pathways to...

75 FR 38564 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-02

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations and Fire Protection The ACRS Subcommittee on Plant Operations and Fire Protection will hold a meeting on July 29, 2010, at the U.S. NRC Region IV, Texas Health Resources Tower, 612...

Status of JUPITER Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Inoue, T.; Shirakata, K.; Kinjo, K.

To obtain the data necessary for evaluating the nuclear design method of a large-scale fast breeder reactor, criticality tests with a large- scale homogeneous reactor were conducted as part of a joint research program by Japan and the U.S. Analyses of the tests are underway in both countries. The purpose of this paper is to describe the status of this project.

78 FR 79506 - Tennessee Valley Authority: Exemption From Requirements To Revise Combined License Application To...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

..., Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington DC, 20555- 0001; telephone: 301...) application for two units of Westinghouse Electric Company's AP1000 advanced pressurized water reactors to be... Bellefonte Nuclear Plant, Units 3 and 4 (BLN 3&4) COL application on January 28, 2008. On September 29, 2010...

78 FR 79501 - Tennessee Valley Authority, Exemption From the Requirement To Submit an Annual Update to the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-30

..., Maryland 20852. FOR FURTHER INFORMATION CONTACT: Anthony Minarik, Office of New Reactors, U.S. Nuclear... advanced pressurized water reactors to be constructed and operated at the Bellefonte site, located near the... 052000-15). The NRC docketed the Bellefonte Nuclear Plant, Units 3 and 4 (BLN 3&4) COL application on...

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

76 FR 26320 - Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance of Amendment to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-06

... Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone (301..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-11107 Filed 5-5-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0071; Docket No. 50-382] Entergy Operations, Inc...

76 FR 13676 - Exelon Generation Company, LLC; Notice of Withdrawal of Application for Amendment to Facility...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-14

.... Brown, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-5756 Filed 3-11-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-373 and 50-374; NRC-2011-0051] Exelon Generation...

75 FR 57535 - Northern States Power Company-Minnesota Notice of Issuance of Amendments to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-21

... issuance. FOR FURTHER INFORMATION CONTACT: Thomas J. Wengert, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2010-23516 Filed 9-20-10; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0290] Northern States Power...

78 FR 17450 - Entergy Nuclear Operations, Inc.; Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-21

... INFORMATION CONTACT: Nageswaran Kalyanam, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2013-06510 Filed 3-20-13; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0012] [Docket Nos. 50-458, 50-155, 72-043, 50-003, 50-247...

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-05-14

fuel for future civilian light water reactors deployed” in the UAE. The agreement also states that future cooperation may encompass training...planned nuclear reactor . (...continued) May 4, 2008; and, Chris Stanton and Ivan...already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and advisory services contracts related to the establishment

Las historias de la narrativa hispanoamericana: Criterios, metodos y ausencias. (Histories of the Latin-American Narrative: Criteria, Methods, and Absences).

ERIC Educational Resources Information Center

Zavalo, Lauro

This paper explains that materials on the teaching of Latin-American literature are sparse, even though most researchers in the field will dedicate much of their time to teaching. The paper adds that, in scholarly journals, little attention is given to teaching literature, and the topic is also absent from most academic congresses. The paper then…

Program for Calculating the Cubic and Fifth Roots of a Number of Newton's Method (610 IBM Electronic Computer); PROGRAMMA PER IL CALCOLO DELLE RADICI TERZA E QUINTA DI UN NUMERO, COL METODO DI NEWTON (CALCOLATORE ELETTRONICO IBM 610)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Giucci, D.

1963-01-01

A program was devised for calculating the cubic and fifth roots of a number of Newton's method using the 610 IBM electronic computer. For convenience a program was added for obtaining n/sup th/ roots by the logarithmic method. (auth)

Estudio comparativo de los metodos analitico-sintetico y global en el aprendizaje de la lectura (Comparative Study of the Analytical-Synthetical (Phonics) and Global (Sight) Reading Methods).

ERIC Educational Resources Information Center

Carbonell de Grompone, Maria A.; And Others

An investigation into the phonics and sight methods of reading instruction being taught in Uruguay schools seeks valid predictions in support of each approach. The study, written in Spanish, examines the progressive reading habits and abilities of 12 first-grade classes. Teachers assigned to teach each method uniformly had equivalent training and…

Metodo y Proceso de la Investigacion Participativa en la Capacitacion Rural (The Method and Process of Participatory Research in Rural Leadership Training). Cuadernos del CREFAL 19.

ERIC Educational Resources Information Center

de Schutter, Anton

In participatory research, education and learner participation are directly connected. The document analyzes the role of a participatory research method in the basic education of rural adults. The different phases of the Participant Research method are presented, along with a profound analysis of both research and participation. The claim is that…

Atlas de aves: Un metodo para documentar distribucion y seguir poblaciones

USGS Publications Warehouse

Robbins, C.S.; Dowell, B.A.; Dawson, D.K.; Alvarez-Lopez, Humberto; Kattan, Gustavo; Murcia, Carolina

1988-01-01

Los Atlas de Aves son proyectos nacionales o regionalies para trazar en mapas la distribucion en reproduccion de cada especie de ave. Ese procedimiento se esta usando en Europa, Australia, Nueva Zelanda, Norteamerica, y partes de Africa. El tama?o de los cuadrados varia de medio grado de latitud y Iongitud hasta 5 x 5 km. El trabajo de campo de cada proyecto exige aproxlmadamente cinco a?os, pero los aficionados pueden llevar a cabo la mayor parte del trabajo. Es posible almacenar los resultados en un computador personal. Hay muchos beneficios: (I) se presenta la distribucion corriente de las aves de la nacion, del estado, o de la Iocalidad; (2) se desarrolla nueva informacion especialmente sobre especies raras o en peligro; (3) se descubren areas que tienen una avlfauna sobresaliente o habitats raros y ayuda a su proteccion, (4) se documentan cambios de dlstribucion; (5) se pueden usar para documentar cambios de poblacion, especialmente en los tropicos donde otros metodos son mas dificiles de usar porque hay muchas especies y no hay muchos observadores calificados en la identificacion de sonidos de las aves; (6) son proyectos buenos de investigacion para estudiantes graduados; (7) los turistas y los jefes de excursiones de historia natural pueden contribuir con muchas informaciones

Waste Generated from LMR-AMTEC Reactor Concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hasan, Ahmed; Mohamed, Yasser, T.; Mohammaden, Tarek, F.

2003-02-25

The candidate Liquid Metal Reactor-Alkali Metal Thermal -to- Electric Converter (LMR-AMTEC) is considered to be the first reactor that would use pure liquid potassium as a secondary coolant, in which potassium vapor aids in the conversion of thermal energy to electric energy. As with all energy production, the thermal generation of electricity produces wastes. These wastes must be managed in ways which safeguard human health and minimize their impact on the environment. Nuclear power is the only energy industry, which takes full responsibility for all its wastes. Based on the candidate design of the LMR-AMTEC components and the coolant types,more » different wastes will be generated from LMR. These wastes must be classified and characterized according to the U.S. Code of Federal Regulation, CFR. This paper defines the waste generation and waste characterization from LMR-AMTEC and reviews the applicable U.S. regulations that govern waste transportation, treatment, storage and final disposition. The wastes generated from LMR-AMTEC are characterized as: (1) mixed waste which is generated from liquid sodium contaminated by fission products and activated corrosion products; (2) hazardous waste which is generated from liquid potassium contaminated by corrosion products; (3) spent nuclear fuel; and (4) low-level radioactive waste which is generated from the packing materials (e.g. activated carbon in cold trap and purification units). The regulations and management of these wastes are summarized in this paper.« less

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

Total electron acceptor loading and composition affect hexavalent uranium reduction and microbial community structure in a membrane biofilm reactor.

PubMed

Ontiveros-Valencia, Aura; Zhou, Chen; Ilhan, Zehra Esra; de Saint Cyr, Louis Cornette; Krajmalnik-Brown, Rosa; Rittmann, Bruce E

2017-11-15

Molecular microbiology tools (i.e., 16S rDNA gene sequencing) were employed to elucidate changes in the microbial community structure according to the total electron acceptor loading (controlled by influent flow rate and/or medium composition) in a H 2 -based membrane biofilm reactor evaluated for removal of hexavalent uranium. Once nitrate, sulfate, and dissolved oxygen were replaced by U(VI) and bicarbonate and the total acceptor loading was lowered, slow-growing bacteria capable of reducing U(VI) to U(IV) dominated in the biofilm community: Replacing denitrifying bacteria Rhodocyclales and Burkholderiales were spore-producing Clostridiales and Natranaerobiales. Though potentially competing for electrons with U(VI) reducers, homo-acetogens helped attain steady U(VI) reduction, while methanogenesis inhibited U(VI) reduction. U(VI) reduction was reinstated through suppression of methanogenesis by addition of bromoethanesulfonate or by competition from SRB when sulfate was re-introduced. Predictive metagenome analysis further points out community changes in response to alterations in the electron-acceptor loading: Sporulation and homo-acetogenesis were critical factors for strengthening stable microbial U(VI) reduction. This study documents that sporulation was important to long-term U(VI) reduction, whether or not microorganisms that carry out U(VI) reduction mediated by cytochrome c 3 , such as SRB and ferric-iron-reducers, were inhibited. Copyright © 2017 Elsevier Ltd. All rights reserved.

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

NASA Astrophysics Data System (ADS)

Miyake, Yasuto; Matsuzaki, Hiroyuki; Sasa, Kimikazu; Takahashi, Tsutomu

2015-10-01

In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as 129I (T1/2 = 1.57 × 107 year) and 36Cl (T1/2 = 3.01 × 105 year). 129I and 131I are both produced by 235U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes 129I useful for retrospective reconstruction of 131I distribution during the initial stages of the accident. On the other hand, 36Cl is generated during reactor operation via neutron capture reaction of 35Cl, an impurity in the coolant or reactor component. Resulting 36Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to 129I, 36Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, 129I concentrations were determined in several surface soil samples collected around F1NPP. Average 129I/131I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. 36Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10-12 to 2.6 × 10-11. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between 36Cl and 129I concentration was observed.

The RERTR Program : a status report.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident conditions. A major milestone was accomplished in the development of a process to produce molybdenum-99 from fission targets utilizing LEU instead of HEU. Targets containing LEU metal foils were irradiated in the RAS-GAS reactor at BATAN, Indonesia, and molybdenum-99 was successfully extracted through the ensuing process. These are exciting times for the program and for all those involved in it, and last year's successes augur well for the future. However, as in the past, the success of the RERTR program will depend on the international friendship and cooperation that have always been its trademark.« less

Advanced applications of cosmic-ray muon radiography

NASA Astrophysics Data System (ADS)

Perry, John

The passage of cosmic-ray muons through matter is dominated by the Coulomb interaction with electrons and atomic nuclei. The muon's interaction with electrons leads to continuous energy loss and stopping through the process of ionization. The muon's interaction with nuclei leads to angular diffusion. If a muon stops in matter, other processes unfold, as discussed in more detail below. These interactions provide the basis for advanced applications of cosmic-ray muon radiography discussed here, specifically: 1) imaging a nuclear reactor with near horizontal muons, and 2) identifying materials through the analysis of radiation lengths weighted by density and secondary signals that are induced by cosmic-ray muon trajectories. We have imaged a nuclear reactor, type AGN-201m, at the University of New Mexico, using data measured with a particle tracker built from a set of sealed drift tubes, the Mini Muon Tracker (MMT). Geant4 simulations were compared to the data for verification and validation. In both the data and simulation, we can identify regions of interest in the reactor including the core, moderator, and shield. This study reinforces our claims for using muon tomography to image reactors following an accident. Warhead and special nuclear materials (SNM) imaging is an important thrust for treaty verification and national security purposes. The differentiation of SNM from other materials, such as iron and aluminum, is useful for these applications. Several techniques were developed for material identification using cosmic-ray muons. These techniques include: 1) identifying the radiation length weighted by density of an object and 2) measuring the signals that can indicate the presence of fission and chain reactions. By combining the radiographic images created by tracking muons through a target plane with the additional fission neutron and gamma signature, we are able to locate regions that are fissionable from a single side. The following materials were imaged with this technique: aluminum, concrete, steel, lead, and uranium. Provided that there is sufficient mass, U-235 could be differentiated from U-238 through muon induced fission.

76 FR 11525 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-02

... the U.S. EPR Document Control Design (DCD) Safety Evaluation Report (SER) with Open Items. The... from the website cited above or by contacting the identified DFO. Moreover, in view of the possibility...

Pore growth in U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D.-S.; Jamison, L. M.

2016-09-01

U-Mo/Al dispersion fuel is currently under development in the DOE's Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

I. Glagolenko; D. Wachs; N. Woolstenhulme

2010-10-01

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 duemore » 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.« less

Expert judgments about RD&D and the future of nuclear energy.

PubMed

Anadón, Laura D; Bosetti, Valentina; Bunn, Matthew; Catenacci, Michela; Lee, Audrey

2012-11-06

Probabilistic estimates of the cost and performance of future nuclear energy systems under different scenarios of government research, development, and demonstration (RD&D) spending were obtained from 30 U.S. and 30 European nuclear technology experts. We used a novel elicitation approach which combined individual and group elicitation. With no change from current RD&D funding levels, experts on average expected current (Gen. III/III+) designs to be somewhat more expensive in 2030 than they were in 2010, and they expected the next generation of designs (Gen. IV) to be more expensive still as of 2030. Projected costs of proposed small modular reactors (SMRs) were similar to those of Gen. IV systems. The experts almost unanimously recommended large increases in government support for nuclear RD&D (generally 2-3 times current spending). The majority expected that such RD&D would have only a modest effect on cost, but would improve performance in other areas, such as safety, waste management, and uranium resource utilization. The U.S. and E.U. experts were in relative agreement regarding how government RD&D funds should be allocated, placing particular focus on very high temperature reactors, sodium-cooled fast reactors, fuels and materials, and fuel cycle technologies.

Inclusion free cadmium zinc tellurium and cadmium tellurium crystals and associated growth method

DOEpatents

Bolotnikov, Aleskey E [South Setauket, NY; James, Ralph B [Ridge, NY

2010-07-20

The present disclosure provides systems and methods for crystal growth of cadmium zinc tellurium (CZT) and cadmium tellurium (CdTe) crystals with an inverted growth reactor chamber. The inverted growth reactor chamber enables growth of single, large, high purity CZT and CdTe crystals that can be used, for example, in X-ray and gamma detection, substrates for infrared detectors, or the like. The inverted growth reactor chamber enables reductions in the presence of Te inclusions, which are recognized as an important limiting factor in using CZT or CdTe as radiation detectors. The inverted growth reactor chamber can be utilized with existing crystal growth techniques such as the Bridgman crystal growth mechanism and the like. In an exemplary embodiment, the inverted growth reactor chamber is a U-shaped ampoule.

A Nuclear Energy Renaissance in the U.S.?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessler, Carol E.; Mahy, Heidi A.; Ankrum, Al

2008-01-01

Is it time for a nuclear energy renaissance? Among other things, nuclear power is a carbon neutral source of base load power. With the growth in energy use expected over the next 20 years and the growing negative impacts of global climate changes, the cost of oil and gas, energy security and diversity concerns, and progress on advanced reactor designs, it may be the right time for nuclear power to enter a new age of growth. Asia and Russia are both planning for a nuclear renaissance. In Europe, Finland and France have both taken steps to pursue new nuclear reactors.more » U.S. utilities are preparing for orders of new reactors; one submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to review its request to construct a new reactor on an existing site. What has the industry been doing since nuclear energy was birthed in the 1960s? In those days a bold new industry boasted that nuclear power in the United States was going to be “too cheap to meter”, but as we all know this did not come about for many reasons. Eventually, it became clear that industry had neglected to do its homework. Critiques of the industry were made on safety, security, environment, economic competitiveness (without government support), and nonproliferation. All of these factors need to be effectively addressed to promote the confidence and support of the public – without which a nuclear power program is not feasible.« less

Dosimetry characterization of the Godiva Reactor under burst conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hickman, D. P.; Heinrichs, D. P.; Hudson, R.

2017-06-22

A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C,more » and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.« less

Nuclear Power Now and in the Near Future

NASA Astrophysics Data System (ADS)

Burchill, William

2006-04-01

The presentation will describe the present status of nuclear power in the United States including its operating, economic, and safety record. This status report will be based on publicly-available records of the U.S. Department of Energy, the U.S. Nuclear Regulatory Commission, and the Institute of Nuclear Power Operations. The report will provide a brief description and state the impact of both the Three Mile Island and Chernobyl accidents. It will list the lessons learned and report significant improvements in U.S. nuclear power plants. The major design differences between Chernobyl and U.S. nuclear reactors will be discussed. The presentation will project the near future of nuclear power considering the 2005 Energy Bill, initiatives by the U.S. Department of Energy and industry, and public opinions. Issues to be considered include plant operating safety, disposition of nuclear waste, protection against proliferation of potential weapons materials, economic performance, environmental impact and protection, and advanced nuclear reactor designs and fuel cycle options. The risk of nuclear power plant operations will be compared to risks presented by other industrial activities.

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise requiredmore » to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.« less

Hydrolysis of lactose by beta-glycosidase CelB from hyperthermophilic archaeon Pyrococcus furiosus: comparison of hollow-fiber membrane and packed-bed immobilized enzyme reactors for continuous processing of ultrahigh temperature-treated skim milk.

PubMed

Splechtna, Barbara; Petzelbauer, Inge; Kuhn, Bernhard; Kulbe, Klaus D; Nidetzky, Bernd

2002-01-01

Recombinant beta-glycosidase CelB from the hyperthermophilic archaeon Pyrococcusfuriosus was produced through expression of the plasmid-encoded gene in Escherichia coli. Bioreactor cultivations of E. coli in the presence of the inductor isopropyl-1-thio-beta-D-galactoside (0.1 mM) gave approx 100,000 U of enzyme activity/L of culture medium after 8 h of growth. A technical-grade enzyme for the hydrolysis of lactose was prepared by precipitating the mesophilic protein at 80 degrees C. A hollow-fiber membrane reactor was developed, and its performance during continuous processing of ultrahigh temperature-treated (UHT) skim milk at 70 degrees C was analyzed regarding long-term stability, productivity, and diffusional limitation thereof. CelB was covalently attached onto Eupergit C in yields of 80%, and a packed-bed immobilized enzyme reactor was used for the continuous hydrolysis of lactose in UHT skim milk at 70 degrees C. The packed-bed reactor was approximately 10-fold more stable and gave about the same productivity at 80% substrate conversion as the hollow-fiber reactor at 60% substrate conversion. The marked difference in the stability of free and immobilized CelB seems to reflect mainly binding of the soluble enzyme to the membrane surface of the hollow-fiber module. Under these bound conditions, CelB is essentially inactive. CelB is essentially inactive. Microbial contamination of the reactors did not occur during reaction times of up to 39 d, given that UHT skim milk and not pasteurized skim milk was used as the substrate.

PBF Reactor Building (PER620). Plot plan shows layout, including auxiliary ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Plot plan shows layout, including auxiliary buildings: Emergency Generator (621), Hose House (622), Cooling Tower Auxiliary (624), Maintenance and Storage Warehouse (625), Gas Cylinder Storage (627), Hose House (628), Cooling Tower (720), Substation (719), and other features. Road connections between PBF Reactor, its control building, and SPERT-I site. Note cable trenches along road to control building. Date: July 1965. Ebasco Services, PER-U-101. INEEL index no. 761-0100-00-205-123005 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Quality control method to measure predator evasion in wild and mass-reared Mediterranean fruit flies (Diptera: Tephritidae)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hendrichs, M.; Wornoayporn, V.; Hendrichs, J.

Sterile male insects, mass-reared and released as part of sterile insect technique (SIT) programs, must survive long enough in the field to mature sexually and compete effectively with wild males for wild females. An often reported problem in Mediterranean fruit fly (medfly) Ceratitis capitata (Wiedemann) SIT programs is that numbers of released sterile males decrease rapidly in the field for various reasons, including losses to different types of predators. This is a serious issue in view that most operational programs release sterile flies at an age when they are still immature. Previous field and field-cage tests have confirmed that fliesmore » of laboratory strains are less able to evade predators than wild flies. Such tests involve, however, considerable manipulation and observation of predators and are therefore not suitable for routine measurements of predator evasion. Here we describe a simple quality control method with aspirators to measure agility in medflies and show that this parameter is related to the capacity of flies to evade predators. Although further standardization of the test is necessary to allow more accurate inter-strain comparisons, results confirm the relevance of measuring predator evasion in mass-reared medfly strains. Besides being a measure of this sterile male quality parameter, the described method could be used for the systematic selection of strains with a higher capacity for predator evasion. (author) [Spanish] Insectos machos esteriles criados en forma masiva para ser liberados en programas que utilizan la tecnica del insecto esteril (TIE), tienen que tener la capacidad de sobrevivir en el campo el tiempo necesario para poder madurar sexualmente y competir efectivamente con los machos silvestres por hembras silvestres. Un problema frecuentemente reportado por dichos programas de la mosca del Mediterraneo, Ceratitis capitata (Wiedemann), es que el numero de machos esteriles de laboratorio liberados en el campo, decrecen rapidamente por varias razones, incluyendo perdidas debidas a diferentes tipos de depredadores. Estudios anteriores conducidos en el campo, y en jaulas de campo, han confirmado que las cepas de machos de laboratorio tienen menos capacidad de evadir depredadores que los machos silvestres. Estos estudios involucran, sin embargo, una considerable cantidad de manipulacion y observacion de depredadores, por lo que no son adecuados para ser usados como medidas rutinarias en los programas de cria masiva. Aqui describimos un metodo sencillo de control de calidad usando aspiradores para medir agilidad en la mosca del Mediterraneo y mostramos que este parametro esta relacionado a la capacidad de la moscas a evadir a depredadores. Aunque aun es necesario refinar la estandarizacion de este metodo para permitir la comparacion entre cepas, los resultados confirman la importancia de tener un metodo rutinario para medir la capacidad de evasion de depredadores en cepas de cria de laboratorio de la mosca del Mediterraneo. Ademas de medir este parametro de control de calidad de los machos esteriles, el metodo descrito podria tambien ser usado para la seleccion sistematica de cepas con una mayor capacidad de evasion de depredadores. (author)« less

Effect of giant charge-transfer resonance σCT 109 barn on operation of magnetic fusion reactor below ``critical energy.''

NASA Astrophysics Data System (ADS)

Hester, Timothy; Maglich, Bogdan; Scott, Dan; Vaucher, Alexander

2016-10-01

Charge transfer (CT) reactivity was assumed to be negligible compared to ionization (IO) before Belfast measurements1-3 revealed the opposite: CT predominance over IO, σCT 109 b , σCT /σIO U 100 , below critical `atomic unit of velocity', vo = 2.2 ×108cms-1 , which is orbital velocity of e in H atom. Near vo, U = 1 , i.e. σCT σIO . Critical ion energy is T0 (lab) = k 25 M [ KeV ] = 200 KeV for [ ERR : md : MbegChr = 0 x 2329 , MendChr = 0 x 232 A , nParams = 1 ] = ion mass [ amu ] = 4 for DT mix ; k = 2 . ``Burnout'' pumping that requires U << 1 is inoperable in the U >> 1 regime whereas CT continually acts like compressor increasing operating vacuum pressure during neutral beam discharge to 10-3 Torr/0.3 s; this, in turn, sets upper limits to ion life-time against neutralization to τ =10-6 s. τ is 105 times shorter than thermalization time constant; hence plasma cannot be created. Lawson4 was unaware of CT resonance; his ``critical temperature'' (30 KeV for DT) should be replaced with T0.

PLUTONIUM-URANIUM ALLOY

DOEpatents

Coffinberry, A.S.; Schonfeld, F.W.

1959-09-01

Pu-U-Fe and Pu-U-Co alloys suitable for use as fuel elements tn fast breeder reactors are described. The advantages of these alloys are ease of fabrication without microcracks, good corrosion restatance, and good resistance to radiation damage. These advantages are secured by limitation of the zeta phase of plutonium in favor of a tetragonal crystal structure of the U/sub 6/Mn type.

75 FR 80854 - Toshiba Corporation; Acceptance for Docketing of an Application for Renewal of the U.S. Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-23

... NUCLEAR REGULATORY COMMISSION [Docket No. 52-044; NRC-2010-0361] Toshiba Corporation; Acceptance for Docketing of an Application for Renewal of the U.S. Advanced Boiling Water Reactor Design Certification On November 2, 2010, Toshiba Corporation (Toshiba) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for a design...

76 FR 59392 - Notice of Intent To Grant Exclusive Patent License; Enhanced Energy Group, LLC

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-26

... inventions, and they are covered by U.S. Patent No. 7,926,275: Closed Brayton Cycle Direct Contact Reactor/ Storage Tank With Chemical Scrubber.//U.S. Patent No. 7,926,276: Closed Cycle Brayton Propulsion System With Direct Heat Transfer.//U.S. Patent No. 7,937,930: Semiclosed Brayton Cycle Power System With...

76 FR 59391 - Notice of Availability of Government-Owned Inventions; Available for Licensing

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-26

... Issued 4/12/2011//U.S. Patent No. 7,926,275: Closed Brayton Cycle Direct Contact Reactor/Storage Tank... With Precursor Bubble Issued 4/19/2011//U.S. Patent No. 7,929,375: Method and Apparatus for Improved...: Hydrogen Generator Apparatus for an Underwater Vehicle Issued 5/10/2011//U.S. Patent No. 7,940,602...

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-10-28

India in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor Export Deal,” Chennai Business Line...MW thermal or special nuclear material connected therewith. U.S. Nuclear Cooperation with India: Issues for Congress Congressional Research

Nuclear Fuels.

ERIC Educational Resources Information Center

Nash, J. Thomas

1983-01-01

Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

U.S. Department of Energy Office of Nuclear Technology Research and Eevelopment ((NTRD) comprehensive summary of QA assessments for FY17

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trost, Alan L.

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) has developed a research and development (R&D) roadmap for its research, development, and demonstration (RD&D) activities to ensure nuclear energy remains a compelling and viable energy option for the U.S. The roadmap defines NE RD&D activities and objectives that address the challenges to research, develop and demonstrate options to the current U.S commercial fuel cycle to enable the safe, secure, economic, and sustainable expansion of nuclear energy, while minimizing proliferation and terrorism risks expanding the use of nuclear power. The roadmap enables the development of technologies and other solutionsmore » that can improve the reliability, sustain the safety, and extend the life of current reactors. In addition, it will help to develop improvements in the affordability of the new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals.« less