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Sample records for standard heu core

  1. Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

    International Nuclear Information System (INIS)

    Pruimboom, H.; Tas, A.

    1985-01-01

    Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

  2. The calculation of the MEU-HEU coupled core in the KUCA

    International Nuclear Information System (INIS)

    Hayashi, M.; Shiroya, S.; Kanda, K.; Shibata, T.

    1984-01-01

    The KUCA has a plan for critical experiments of the MEU-HEU coupled core in 1984. The neutronics calculation has been performed for the MEU-HEU coupled core in the KUCA. The GGC-4 and THERMOS were used to generate the four-group constants and the 2D-FEM-KUR, based on the finite-element method, was used for the diffusion calculation. The calculations with four-group constants agreed with experiments within 1.8% for the both single-cores with the MEU and the HEU. (author)

  3. Neutronic analysis of the conversion of HEU to LEU fuel for a 5-MW MTR core

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Bartsch, G.

    1987-01-01

    In recent years, due to cessation of highly enriched uranium (HEU) fuel supply, practical steps have been taken to substitute HEU fuel in almost all research reactors by medium-enriched uranium or low-enriched uranium (LEU) fuels. In this study, a neutronic calculation of a 5-MW research reactor core fueled with HEU (93% 235 U) is presented. In order to assess the performance of the core with the LEU ( 235 U loadings were examined. The core consists of 22 standard fuel elements (SFEs) and 6 control fuel elements (CFEs). Each fuel elements has 18 curved plates of which two end plates are dummies. Initial 235 U content is 195 g 235 U/SFE and 9.7 g 235 U/CFE or /PFE. In all calculations the permitted changes to the fuel elements are (a) 18 active plates per SFE, (b) fuel plates assumed to be flat, and (c) 8 or 9 active plates per CFE

  4. HEU/LEU-conversion of BER II successfully finished

    International Nuclear Information System (INIS)

    Haas, K.; Fischer, C.-O.; Krohn, H.

    2000-01-01

    The BER II (Berliner Experimental Reactor) research reactor is a swimming pool type reactor located in Berlin, Germany. The reactor operates with a thermal power of 10 MW and is primarily used to produce neutrons for neutron scattering experiments. The conversion from HEU- to LEU-fuel elements began in August, 1997. At the last RERTR Meeting 1999 in Budapest, Hungary, Hahn-Meitner-Institut (HMI) presented a 'Status Report' on the conversion of 10 HEU/LEU mixed cores. In February 2000, HMI finished the HEU/LEU-conversion. Hereby, the first pure LEU-standard-core went into operation. Our second LEU-core just ends its operation at the end of July. The third LEU-core will be built up in the beginning of August. The average burn-up rate was improved from 50 - 55% (HEU) to 60 - 65% (LEU). Therefore, only 14 elements/year are now used instead of 28/year. The following report describes our first steps in building pure LEU-cores from mixed HEU/LEU-cores, as well as our initial experience using the pure LEU-cores. (author)

  5. Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

    International Nuclear Information System (INIS)

    Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

    1997-01-01

    The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

  6. Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz

    2000-01-01

    This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

  7. Mixed core conversion study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1984-01-01

    The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically

  8. Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1985-07-01

    The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

  9. Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Nabbi, R.

    1987-02-01

    In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

  10. A mixed core conversion study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1985-01-01

    The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically. (author)

  11. The status of HEU and LEU core conversion activities at the Jamaica SLOWPOKE

    Energy Technology Data Exchange (ETDEWEB)

    Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Kingston (Jamaica)

    2013-07-01

    The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, inline with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

  12. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  13. The status of HEU to LEU core conversion activities at the Jamaica SLOWPOKE

    Energy Technology Data Exchange (ETDEWEB)

    Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Mona (Jamaica)

    2012-12-15

    The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, in line with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

  14. Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

    International Nuclear Information System (INIS)

    Arigane, Kenji

    1987-04-01

    The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

  15. Studies of mixed HEU-LEU-MTR cores using 3D models

    Energy Technology Data Exchange (ETDEWEB)

    Haenggi, P.; Lehmann, E.; Hammer, J.; Christen, R. [Paul Scherrer Institute, Villigen (Switzerland)

    1997-08-01

    Several different core loadings were assembled at the SAPHIR research reactor in Switzerland combining the available types of MTR-type fuel elements, consisting mainly of both HEU and LEU fuel. Bearing in mind the well known problems which can occur in such configurations (especially power peaking), investigations have been carried out for each new loading with a 2D neutron transport code (BOXER). The axial effects were approximated by a global buckling value and therefore the radial effects could be studied in considerably detail. Some of the results were reported at earlier RERTR meetings and were compared to those obtained by other methods and with experimental values. For the explicit study of the third dimension of the core, another code (SILWER), which has been developed in PSI for LWR power plant cores, has been selected. With the help of an adapted model for the MTR-core of SAPHIR, several important questions have been addressed. Among other aspects, the estimation of the axial contribution to the hot channel factors, the influence of the control rod position and of the Xe-poisoning on the power distribution were studied. Special attention was given to a core position where a new element was assumed placed near a empty, water filled position. The comparison of elements of low and high enrichments at this position was made in terms of the induced power peaks, with explicit consideration of axial effects. The program SILWER has proven to be applicable to MTR-cores for the investigation of axial effects. For routine use as for the support of reactor operation, this 3D code is a good supplement to the standard 2D model.

  16. Determination of Dancoff correction thermal utilization and thermal disadvantage factors of HEU and LEU cores of an MNSR

    International Nuclear Information System (INIS)

    Ofori, Y. T.

    2013-07-01

    Ghana Research Reactor-1 (GHARR-1), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (Highly Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of the conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. In this research work, a comparative study has been performed for the determination of the Dancoff, thermal utilization and thermal disadvantage factors of highly enriched uranium (HEU) and potential low enriched uranium (LEU) cores of GHARR-1. A one group transport theory and collision probability based methodologies was used to develop mathematical formulations for thermal utilization factor and thermal disadvantage factor assuming isotropic scattering. This methodology was implemented in a FORTRAN 95 based computer program THERMCALC, which uses Bessell and BesselK as subroutines developed to calculate the modified Bessel functions I n and K n respectively using the polynomial approximation method. Furthermore, a Dancoff correction factor of 0.1519 thermal utilization factor of 0.9767 and a thermal disadvantage factor of 1.894 were obtained for the 90.2% highly enriched Uranium core of GHARR-1. The results compare favorably with literature. Thus THERMCALC can be used as a reliable tool for the calculation of Dancoff, thermal utilization and disadvantage factors of MNSR cores. Other potential LEU cores; UO 2 (with different fuel meat densities and enrichments) and U 3 Si 2 have also been analysed. UO 2 with 12.6% of Uranium-235 was chosen as the most potential LEU core for the GHARR-1. (au)

  17. Fuel conversion of JRR-4 from HEU to LEU

    International Nuclear Information System (INIS)

    Ichikawa, Hiroki; Nakajima, Teruo

    1997-01-01

    Japanese JRR-4 (Japan Research Reactor No.4) is a pool type, light water moderated and cooled, ETR type fuel reactor used for Shielding experiments, isotope production, neutron activation analyses, Si doping, reactor students training. It acieved first criticality on January 28, 1965 with maximum thermal power 3.5MW. The standard core consistes of 20 Fuel elements, 7 control rods 5 Irradiation holes, neutron source, graphite reflectors. Available thermal flux is 7x1013 n/cm2/s. Within the RERTR program plans are made for core conversion from HEU to LEU

  18. 2011 Progress Report on HEU Minimization Activities in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Bonini, A.; Cristini, P.; Lio, L. De; Dell' Occhio, L.; Gil, D.; Gonzalez, A.G.; Gonzalez, R.; Varela, C. Komar; Lopez, M.; Novara, O.; Taboada, H. [Comision Nacional de Energia Atomica, Av. Del Libertador 8250 (1429) Buenos Aires (Argentina)

    2011-07-01

    After the core conversion of the RA-6 reactor finished in March 2008, an extension of the original CNEA-NNSA DoE contract was signed to enhance the final national HEU inventories minimization. Before this process, CNEA reserved a small inventory of HEU for R and D uses in fission chambers, neutronic probes and standards. This minimization comprises that all fresh and irradiated HEU remnant inventories coming from fuels and Mo99 irradiation targets fabrication and irradiated HEU-oxides retained in production filters and solutions will be recovered, down-blended into LEU and purified or dispose as waste whenever its recovery would not be advisable due to cost-benefit consideration. CNEA has a R and D program to develop the fabrication technology of both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to support the qualification activities of the RERTR program. Some monolithic 58% enrichment and LEU 8%Mo and U10%Mo miniplates and plates were and are being delivered to INL-DoE to be irradiated in the ATR reactor core. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with 1/3 of the national requirements on Mo99 by weekly deliveries. Australia has started the fission radioisotope production through several batches by week, based on CNEA's LEU technology provided by INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1. Plans to recover and purify the LEU based inventories in Mo99 production filters, once the HEU to LEU campaign is over. 2. Fabrication and delivering to INL to be irradiated in the ATR core of U-8%Mo and U-10%Mo monolithic miniplates and development and fabrication of LEU very high density monolithic and dispersed U-Mo fuel plates with Zr cladding for the FUTURE-MONO experiment in the frame of the RERTR program. 3

  19. HEU core conversion of Russian production reactors: a major threat to the international RERTR regime

    International Nuclear Information System (INIS)

    Kuperman, Alan J.; Leventhal, Paul L.

    1998-01-01

    This paper calls the attention for the major threat to the International Reduced Enrichment for Research and Test Reactors (RERTR) program, represented by the HEU core conversion of russian production reactors. This program aims to reduce and eventually eliminate international civilian commerce in nuclear weapons-usable, highly enriched uranium , and thereby significantly lower risks of the material being stolen or diverted by terrorist or states for producing nuclear weapons

  20. Comparison of the FRM-II HEU design with an alternative LEU design

    International Nuclear Information System (INIS)

    Mo, S.C.; Hanan, N.A.; Matos, J.E.

    2004-01-01

    The FRM-II reactor design of the Technical University of Munich has a compact core that utilizes fuel plates containing highly-enriched uranium (HEU, 93%). This paper presents an alternative core design utilizing low-enriched uranium (LEU, 3 that provides nearly the same neutron flux for experiments as the HEU design, but has a less favourable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 - 6.5 g/cm 3 . were developed, the alternative design would provide the same neutron flux and use the same number of cores per year as the HEU design. The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. Further optimization of the LEU design and near-term availability of LEU fuel with a uranium density greater than 4.8 g/cm 3 would enhance the performance of the LEU core. The REKIR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance utilizing LEU fuel. (author)

  1. A re-evaluation of physical protection standards for irradiated HEU fuel

    International Nuclear Information System (INIS)

    Lyman, Edwin; Kuperman, Alan

    2002-01-01

    In the post-September 11 era, it is essential to reconsider all the assumptions upon which the physical protection systems of the past were based and determine whether these assumptions are still appropriate in light of the current terrorist threat. For instance, the U.S. Nuclear Regulatory Commission definition of a 'formula quantity' of special nuclear material is derived from the belief that a terrorist plot to carry out multiple coordinated attacks on different facilities with the goal of acquiring enough SNM for a nuclear weapon is incredible. This assumption has clearly been proven wrong by the September 11 attacks. Another standard that needs to be revisited is the 'self-protection' threshold that determines whether or not an item containing SNM is considered to be 'irradiated' for physical protection purposes. The current value of this threshold, 1 Sv/hr unshielded at 1 meter, is of questionable value as a deterrent to determined terrorists who would be willing to sustain long-term injury as long as they could accomplish their near-term goals. A more credible threshold would be set at a level that would have a high likelihood of disabling the perpetrators before they could complete their mission. Most irradiated nonpower reactor fuels would be unable to meet such a standard. This raises serious questions about the adequacy of the level of physical protection applied today to the large inventories of irradiated HEU fuels now scattered in storage sites around the world. The absence of a coherent global policy for dealing with these materials has created a situation rife with vulnerabilities that terrorists could exploit. The international community, now seized with concern about unused stockpiles of unirradiated HEU fuels around the world, also needs to appreciate the dangers posed by lightly irradiated spent fuels as well. A U.S. proposal to import Russian HEU for supplying U.S. nonpower reactors will only prolong this situation This paper will review policy

  2. Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

    International Nuclear Information System (INIS)

    Sarta Fuentes, Jose A.; Castiblanco, L.A.

    2003-01-01

    With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)

  3. Status of HEU-LEU conversion of FRJ-2

    International Nuclear Information System (INIS)

    Damm, G.; Nabbi, R.

    2002-01-01

    The operator of the German FRJ-2 research reactor, 'Research Center Juelich', has participated from the beginning in the RERTR programme and made comprehensive contributions to the test and use of LEU fuel for HEU-LEU-conversion measures. The originally planned time scale for the conversion of FRJ-2 was significantly delayed because of a change of the manufacturer of the LEU fuel elements and a 4 years shutdown of the reactor for refurbishment purposes. In the meantime the new LEU fuel elements are qualified and tested in the reactor. In the moment calculations for the safety report are made and it is planned to apply for the license of FRJ-2 operation with LEU fuel at the beginning of 2003. In order to get most reliable results a sophisticated computational method based on a MCNP model coupled with the depletion code BURN was developed for reactor physical calculations, core conversion studies and fuel element performance analysis and applied to the mixed and LEU core. The licensing schedule and results of latest calculations for the conversion study will be presented. The simulations shows that the thermal flux in the LEU core is about 19% resulting in a lower burnup rate. But in the reflector area around the core and in the center of the cold n source the neutron flux reduction remains limited to 6%. Due to a harder neutron spectrum in the LEU core the kinetic and safety related parameters are slightly reduced. Using the ORIGEN code it could be shown that the increase of the total fission products inventory amounts to about 6% compared to a HEU core. As a consequence of the high amount of U-238, the amount of U-235 in the LEU core has to be about 27% higher than in the HEU core but the U-235 burnup is approx. 5% lower due to the contribution of fissile plutonium. (author)

  4. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

    International Nuclear Information System (INIS)

    Schickler, R.A.; Marcum, W.R.; Reese, S.R.

    2013-01-01

    Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

  5. Highly enriched uranium (HEU) storage and disposition program plan

    International Nuclear Information System (INIS)

    Arms, W.M.; Everitt, D.A.; O'Dell, C.L.

    1995-01-01

    Recent changes in international relations and other changes in national priorities have profoundly affected the management of weapons-usable fissile materials within the United States (US). The nuclear weapon stockpile reductions agreed to by the US and Russia have reduced the national security requirements for these fissile materials. National policies outlined by the US President seek to prevent the accumulation of nuclear weapon stockpiles of plutonium (Pu) and HEU, and to ensure that these materials are subjected to the highest standards of safety, security and international accountability. The purpose of the Highly Enriched Uranium (HEU) Storage and Disposition Program Plan is to define and establish a planned approach for storage of all HEU and disposition of surplus HEU in support of the US Department of Energy (DOE) Fissile Material Disposition Program. Elements Of this Plan, which are specific to HEU storage and disposition, include program requirements, roles and responsibilities, program activities (action plans), milestone schedules, and deliverables

  6. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

    2013-09-15

    Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

  7. Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

    2008-01-01

    The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

  8. HEU to LEU conversion experience at the UMass-Lowell research reactor

    International Nuclear Information System (INIS)

    White, John R.; Bobek, Leo M.

    2005-01-01

    The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

  9. Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

    International Nuclear Information System (INIS)

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

    1995-01-01

    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 x 5 square array of HEU U (10 wt% - ZrH - Er 2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incoloy. 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 of the original 29 fuel clusters had an average 235 U burnup in the range from 50 to 62%. Because of the U.S. 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% 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. (author)

  10. HEU Transparency Implementation Program and its Radiation Safety Program

    International Nuclear Information System (INIS)

    Radev, R

    2002-01-01

    of the agreement are met. The Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP), within NNSA implements the transparency provisions of the bilateral agreement. It is constantly making progress towards meeting its objectives and gathering the information necessary to confirm that Russian weapons-usable HEU is being blended into LEU. Since the first shipment in 1995 through December 2001, a total of 141 MT of weapons-grade HEU, about 28% of the agreed total and equivalent to 5,650 nuclear weapons, was converted to LEU, further reducing the threat of this material returning back into nuclear weapons. In the year 2001, the LEU sold to electric utility customers for fuel was sufficient to supply the annual fuel needs for about 50 percent of the U.S. installed nuclear electrical power generation capacity. There are four primary uranium processing activities involved in converting HEU metal components extracted from dismantled nuclear weapons into fuel for power reactors: (1) Converting HEU metal to purified HEU oxide; (2) Converting purified HEU oxide to HEU hexafluoride; (3) Downblending HEU hexafluoride to LEU hexafluoride; and (4) Converting LEU hexafluoride into reactor fuel. The first three processes are currently being performed at four Russian nuclear processing facilities: Mayak Production Association (MPA), Electrochemical Plant (ECP), Siberian Chemical Enterprise (SChE), and Ural Electrochemical Integrated Plant (UEIP). Following the blending down of HEU, the LEU hexafluoride is loaded into industry, standard 30B cylinders at the downblending facilities and transported to St. Petersburg, Russia. From there the LEU is shipped by sea to the United States where it is converted into fuel to be used in nuclear power plants. There are six U.S. facilities processing LEU subject to the HEU purchase agreement: the Portsmouth uranium enrichment plant, Global Nuclear Fuel -America, Framatome-Lynchburg, Framatome-Richland, Westinghouse-Hematite, and

  11. The ORR Whole-Core LEU Fuel Demonstration

    International Nuclear Information System (INIS)

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

    1990-01-01

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U 3 Si 2 -Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235 U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235 U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs

  12. A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

    Energy Technology Data Exchange (ETDEWEB)

    Moss, R L; May, P

    1985-07-01

    The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

  13. A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

    International Nuclear Information System (INIS)

    Moss, R.L.; May, P.

    1985-01-01

    The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

  14. Stationary and protable instruments for assay of HEU [highly enriched uranium] solids holdup

    International Nuclear Information System (INIS)

    Russo, P.A.; Sprinkle, J.K. Jr.; Stephens, M.M.; Brumfield, T.L.; Gunn, C.S.; Watson, D.R.

    1987-01-01

    Two NaI(Tl)-based instruments, one stationary and one portable, designed for automated assay of highly enriched uranium (HEU) solids holdup, are being evaluated at the scrap recovery facility of the Oak Ridge Y-12 Plant. The stationary instrument, a continuous monitor of HEU within the filters of the chip burner exhaust system, measures the HEU deposits that accumulate erratically and rapidly during chip burner operation. The portable system was built to assay HEU in over 100 m of elevated piping used to transfer UO 3 , UO 2 , and UF 4 powder to, from, and between the fluid bed conversion furnances and the powder storage hoods. Both instruments use two detector heads. Both provide immediate automatic readout of accumulated HEU mass. The 186-keV 235 U gamma ray is the assay signature, and the 60-keV gamma ray from an 241 Am source attached to each detector is used to normalize the 186-keV rate. The measurement geometries were selected for compatibility with simple calibration models. The assay calibrations were calculated from these models and were verified and normalized with measurements of HEU standards built to match geometries of uniform accumulations on the surfaces of the process equipment. This instrumentation effort demonstrates that simple calibration models can often be applied to unique measurement geometries, minimizing the otherwise unreasonable requirements for calibration standards and allowing extension of the measurements to other process locations

  15. Verification experiment on the downblending of high enriched uranium (HEU) at the Portsmouth Gaseous Diffusion Plant. Digital video surveillance of the HEU feed stations

    International Nuclear Information System (INIS)

    Martinez, R.L.; Tolk, K.; Whiting, N.; Castleberry, K.; Lenarduzzi, R.

    1998-01-01

    As part of a Safeguards Agreement between the US and the International Atomic Energy Agency (IAEA), the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, was added to the list of facilities eligible for the application of IAEA safeguards. Currently, the facility is in the process of downblending excess inventory of HEU to low enriched uranium (LEU) from US defense related programs for commercial use. An agreement was reached between the US and the IAEA that would allow the IAEA to conduct an independent verification experiment at the Portsmouth facility, resulting in the confirmation that the HEU was in fact downblended. The experiment provided an opportunity for the DOE laboratories to recommend solutions/measures for new IAEA safeguards applications. One of the measures recommended by Sandia National Laboratories (SNL), and selected by the IAEA, was a digital video surveillance system for monitoring activity at the HEU feed stations. This paper describes the SNL implementation of the digital video system and its integration with the Load Cell Based Weighing System (LCBWS) from Oak Ridge National Laboratory (ORNL). The implementation was based on commercially available technology that also satisfied IAEA criteria for tamper protection and data authentication. The core of the Portsmouth digital video surveillance system was based on two Digital Camera Modules (DMC-14) from Neumann Consultants, Germany

  16. 31 CFR 540.306 - Highly Enriched Uranium (HEU).

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Highly Enriched Uranium (HEU). 540...) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly...

  17. Comparison of the FRM-II HEU design with an alternative LEU design. Attachment

    International Nuclear Information System (INIS)

    Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

    2004-01-01

    After presentation of the foregoing paper by Dr. Nelson Hanan of Argonne National Laboratory (ANL) proposing an alternative LEU core with one fuel ring and a power level of 33 MW, a presentation was made by Dr. Klaus Boning of the Technical University of Munich comparing the FRM-II HEU design with an LEU design by Tlm that had two fuel rings and a power level of 40 MW. Dr. Boning raised the following issues concerning the use of LEU fuel in FRM-H reactor designs: (1) qualification of HEU and LEU silicide fuels, (2) gamma heating in the heavy water reflector, (3) the radiological consequences of hypothetical accidents, and (4) cost and schedule. These issues are addressed in this Attachment. In his presentation, Dr. Hanan mentioned that ANL was also investigating other LEU designs. This work led to a second alternative LEU design that has the same neutron flux performance (8 x 10 14 n/cm 2 /s peak neutron flux in the reflector) and the same fuel lifetime (50 full power days) as the HEU design, but uses LEU silicide fuel with a uranium density of only 4.5 g/cm 3 . This design was achieved by using a fuel plate that has a fuel meat thickness of 0.76 mm, a cladding thickness of 0.38 mm, and a water channel gap of 2.2 mm. A comparison is shown of the main characteristics of this second alternative LEU design with those of the FRM-II HEU design. The ANL core again has one fuel ring with the same dimensions. With this LEU design, a two stage process is no longer necessary because LEU silicide fuel with a uranium density of 4.5 g/cm 3 is fully qualified, licensable, and available now for use in a high flux reactor such as the FRM-II

  18. Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

    International Nuclear Information System (INIS)

    Hanan, N. A.; Deen, J.R.; Matos, J.E.

    2005-01-01

    Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

  19. Documentation Experiences for Jamaican SLOWPOKE-2 Conversion from HEU to LEU

    International Nuclear Information System (INIS)

    Warner, T.-A.; Dennis, H.; Antoine, J.

    2015-01-01

    The Jamaican SLOWPOKE–2 (JM–1) is a 20 kW research reactor manufactured by Atomic Energy of Canada Limited and has been operating since March 1984, in the department of the International Centre for Environmental and Nuclear Sciences (ICENS), at the University of the West Indies, Mona Campus in Kingston, Jamaica. The pool type reactor has been primarily used for Neutron Activation Analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration. The University, assisted by the IAEA under the GTRI/RERTR program, is currently in the process of converting from HEU to LEU. Extensive documentation on policies, general requirements, elements of the conversion quality assurance (QA) system and conversion QA administrative procedures is required for the conversion. The core conversion activities are being carried out in accordance with current international standards and regulatory guidelines of the newly established Jamaican Radiation Safety Authority (RSA) with agreement between the RSA and IAEA or DOE related to Nuclear Safety and Control. The documentation structure has taken into consideration nuclear safety and licensing, LEU fuel design and conversion analysis, LEU fuel procurement and fabrication, removal of HEU fuel and reactor maintenance and conversion and commissioning, with the conversion QA manual at the apex of the structure. To a large extent, the documentation format will adhere to that of the IAEA applicable regulatory standards and guidance documents. The major challenge of the conversion activities, it is envisioned, will come from the absence of any previous regulatory framework in Jamaica; however, a timeline for the process, which includes training and equipping of regulators, will guide operation. (author)

  20. Operating experience, measurements, and analysis of the LEU whole core demonstration at the FNR

    International Nuclear Information System (INIS)

    Weha, D.K.; Drumm, C.R.; King, J.S.; Martin, W.R.; Lee, J.C.

    1984-01-01

    The 2-MW Ford Nuclear Reactor at the University of Michigan is serving as the demonstration reactor for the MTR-type low enrichment (LEU) fuel for the Reduced Enrichment for Research and Test Reactor program. Operational experience gained through six months of LEU core operation and seven months of mixed HEU-LEU core operation is presented. Subcadmium flux measurements performed with rhodium self-powered neutron detectors and iron wire activations are compared with calculations. Measured reactivity parameters are compared for HEU and LEU cores. Finally, the benchmark calculations for several HEU, LEU, and mixed HEU-LEU FNR cores and the International Atomic Energy Agency (IAEA) benchmark problem are presented. (author)

  1. A level-playing field for medical isotope production - How to phase-out reliance on HEU

    International Nuclear Information System (INIS)

    Kuperman, A.J.

    1999-01-01

    Two decades ago, civilian commerce in highly enriched uranium (HEU) for use as targets in the production of medical isotopes was considered a relatively minor security concern for three reasons. First, the number of producers was small. Second, the amount of HEU involved was small. Third, the amount of HEU was dwarfed by the quantities of HEU in civilian commerce as fuel for nuclear research and test reactors. Now, however, all three variables have changed. First, as the use of medical isotopes has expanded rapidly, production programs are proliferating. Second, as the result of such new producers and the expansion of existing production facilities, the amounts of HEU involved are growing. Third, as the RERTR program has facilitated the phase-out of HEU as fuel in most research and test reactors, the quantities of HEU for isotope production have come to represent a significant percentage of global commerce in this weapons-usable material. Medical isotope producers in several states are cooperating with the RERTR program to convert to low-enriched uranium (LEU) targets within the next few years, and one already relies on LEU for isotope production. However, the three biggest isotope producers - in Canada and the European Union - continue to rely on HEU, creating a double-standard that endangers the goal of the RERTR program. Each of these three producers has expressed economic concerns about being put at a competitive disadvantage if it alone converts. This paper proposes forging a firmer international consensus that all present and future isotope producers should convert to LEU, and calls for codifying such a commitment in a statement of intent to be prepared by producers over the next year. With such a level playing field, no producer would need fear being put at a competitive disadvantage by conversion, or being stigmatized by pressure groups for continued reliance on HEU. The phase-out of all HEU commerce for isotope production could be achieved within about

  2. TRANSPARENCY: Tracking Uranium under the U.S./Russian HEU Purchase Agreement

    International Nuclear Information System (INIS)

    Benton, J B; Decman, D J; Leich, D A

    2005-01-01

    By the end of August, 2005, the Russia Federation delivered to the United States (U.S.) more than 7,000 metric tons (MT) of low enriched uranium (LEU) containing approximately 46 million SWU and 75,000 MT of natural uranium. This uranium was blended down from weapons-grade (nominally enriched to 90% 235 U) highly enriched uranium (HEU) under the 1993 HEU Purchase Agreement that provides for the blend down of 500 MT HEU into LEU for use as fuel in commercial nuclear reactors. The HEU Transparency Program, under the National Nuclear Security Administration (NNSA), monitored the conversion and blending of the more than 250 MT HEU used to produce this LEU. The HEU represents more than half of the 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the elimination of more than 10,000 nuclear devices. The HEU Transparency Program has made considerable progress in its mission to develop and implement transparency measures necessary to assure that Russian HEU extracted from dismantled Russian nuclear weapons is blended down into LEU for delivery to the United States. U.S. monitor observations include the inventory of inprocess containers, observation of plant operations, nondestructive assay measurements to determine 235 U enrichment, as well as the examination of Material Control and Accountability (MC and A) documents. During 2005, HEU Transparency Program personnel will conduct 24 Special Monitoring Visits (SMVs) to four Russian uranium processing plants, in addition to staffing a Transparency Monitoring Office (TMO) at one Russian site

  3. Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

  4. 2009 Annual Health Physics Report for the HEU Transparency Program

    International Nuclear Information System (INIS)

    Radev, R.

    2010-01-01

    During the 2009 calendar year, Lawrence Livermore National Laboratory (LLNL) provided health physics support for the Highly Enriched Uranium (HEU) Transparency Program for external and internal radiation protection. LLNL also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2009, there were 159 person-trips that required dose monitoring of the U.S. monitors. Of the 159 person-trips, 149 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 4 monitoring visits by TMO monitors to facilities other than UEIE and 10 to UEIE itself. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2009, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency Program now has over fifteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

  5. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  6. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  7. 2011 Annual Health Physics Report for the HEU transparency Program

    International Nuclear Information System (INIS)

    Radev, R.

    2012-01-01

    During the 2008 calendar year, Lawrence Livermore National Laboratory (LLNL) provided health physics support for the Highly Enriched Uranium (HEU) Transparency Program for external and internal radiation protection. They also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2008, there were 158 person-trips that required dose monitoring of the U.S. monitors. Of the 158 person-trips, 148 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 6 monitoring visits by TMO monitors to facilities other than UEIE and 8 to UEIE itself. There were three monitoring visits (source changes) that were back-to-back with a total of 24 monitors. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2008, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency now has thirteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

  8. Fuel depletion analyses for the HEU core of GHARR-1: Part II: Fission product inventory

    International Nuclear Information System (INIS)

    Anim-Sampong, S.; Akaho, E.H.K.; Boadu, H.O.; Intsiful, J.D.K.; Osae, S.

    1999-01-01

    The fission product isotopic inventories have been estimated for a 90.2% highly enriched uranium (HEU) fuel lattice cell of the Ghana Research Reactor-1 (GHARR-1) using the WIMSD/4 transport lattice code. The results indicate a gradual decrease in the Xe 135 inventory, and saturation trend for Sm 149 , Cs 134 and Cs 135 inventories as the fuel is depleted to 10,000 MWd/tU. (author)

  9. HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Margeanu, S.; Barbos, D.

    2009-01-01

    INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

  10. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

    Waldman, R.M.; Vertullo, A.C.

    1987-01-01

    The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

  11. Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

    International Nuclear Information System (INIS)

    Hernandez G, J.

    2012-10-01

    Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

  12. Analysis of the Ford Nuclear Reactor LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1983-09-01

    This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

  13. Fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1984-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

  14. Energy–angle correlation of neutrons and gamma-rays emitted from an HEU source

    Energy Technology Data Exchange (ETDEWEB)

    Miloshevsky, G., E-mail: gennady@purdue.edu; Hassanein, A.

    2014-06-01

    Special Nuclear Materials (SNM) yield very unique fission signatures, namely correlated neutrons and gamma-rays. A major challenge is not only to detect, but also to rapidly identify and recognize SNM with certainty. Accounting for particle multiplicity and correlations is one of standard ways to detect SNM. However, many parameter data such as joint distributions of energy, angle, lifetime, and multiplicity of neutrons and gamma-rays can lead to better recognition of SNM signatures in the background radiation noise. These joint distributions are not well understood. The Monte Carlo simulations of the transport of neutrons and gamma-rays produced from spontaneous and interrogation-induced fission of SNM are carried out using the developed MONSOL computer code. The energy spectra of neutrons and gamma-rays from a bare Highly Enriched Uranium (HEU) source are investigated. The energy spectrum of gamma-rays shows spectral lines by which HEU isotopes can be identified, while those of neutrons do not show any characteristic lines. The joint probability density function (JPDF) of the energy–angle association of neutrons and gamma-rays is constructed. Marginal probability density functions (MPDFs) of energy and angle are derived from JPDF. A probabilistic model is developed for the analysis of JPDF and MPDFs. This probabilistic model is used to evaluate mean values, standard deviations, covariance and correlation between the energy and angle of neutrons and gamma-rays emitted from the HEU source. For both neutrons and gamma-rays, it is found that the energy–angle variables are only weakly correlated.

  15. RERTR end-game: A win-win framework. Phasing out remaining global HEU commerce by conditionally and temporarily renewing U.S. exports of HEU

    International Nuclear Information System (INIS)

    Kuperman, Alan J.; Leventhal, Paul L.

    1997-01-01

    The RERTR program stands on the brink of fulfilling its historic mission. However, a series of missteps and misunderstandings have recently raised the risk that defeat will be snatched from the jaws of victory. Perhaps the most serious threat to the RERTR regime is posed by France's pending import of 625 kilograms of bomb-grade, highly enriched uranium (HEU) from Russia, intended primarily to fuel its high-flux research reactor at the Institute Laue-Langevin in Grenoble, as well as its Orphee research reactor. As the first export of HEU from Russia to a facility outside the former Soviet bloc, this precedential transaction would establish Russia as a new global supplier of bomb-grade uranium, potentially setting the stage for a rise in international HEU commerce, rather than its phase-out as envisioned under the RERTR program. Apparently, France turned to Russia for supply of the fuel because the United States was perceived as unable or unwilling to continue supplying such fuel in the wake of the U.S. Energy Policy Act of 1992, which, pursuant to its so-called Schumer Amendment, places sharp restrictions on HEU exports. Unexplained delays in Russia's shipment of this material to France provide a fortuitous window of opportunity in which efforts can and should be made by France and the United States to resolve present differences in a manner beneficial to each, as well as in the interest of global security. This paper proposes an arrangement under which the United States would renew exports of HEU to France, in exchange for pledges from France enabling the export to comply with the principles and objectives of the RERTR program as embodied in U.S. law. In so doing, the arrangement would obviate the need for Russian HEU export, thereby avoiding its dangerous precedent. By enabling high quality scientific research to continue, while simultaneously helping to fulfill the RERTR program's original goal, such an arrangement would truly be a 'win-win' solution. (author)

  16. Sealing of process valves for the HEU downblending verification experiment at Portsmouth

    International Nuclear Information System (INIS)

    Baldwin, G.T.; Bartberger, J.C.; Jenkins, C.D.; Perlinski, A.W.; Schoeneman, J.L.; Gordon, D.M.; Whiting, N.E.; Bonner, T.N.; Castle, J.M.

    1998-01-01

    At the Portsmouth Gaseous Diffusion Plant in Piketon, Ohio, USA, excess inventory of highly-enriched uranium (HEU) from US defense programs is being diluted to low-enriched uranium (LEU) for commercial use. The conversion is subject to a Verification Experiment overseen by the International Atomic Energy Agency (IAEA). The Verification Experiment is making use of monitoring technologies developed and installed by several DOE laboratories. One of the measures is a system for sealing valves in the process piping, which secures the path followed by uranium hexafluoride gas (UF 6 ) from cylinders at the feed stations to the blend point, where the HEU is diluted with LEU. The Authenticated Item Monitoring System (AIMS) was the alternative proposed by Sandia National Laboratories that was selected by the IAEA. Approximately 30 valves were sealed by the IAEA using AIMS fiber-optic seals (AFOS). The seals employ single-core plastic fiber rated to 125 C to withstand the high-temperature conditions of the heated piping enclosures at Portsmouth. Each AFOS broadcasts authenticated seal status and state-of-health messages via a tamper-protected radio-frequency transmitter mounted outside of the heated enclosure. The messages are received by two collection stations, operated redundantly

  17. Conversion, core redesign and upgrade of the Rhode Island Atomic Energy Commission Reactor

    International Nuclear Information System (INIS)

    DiMeglio, A.F.

    1987-01-01

    The 2 MW Rhode Island Atomic Energy Commission reactor is required to convert from the use of High Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel using a standard LEU fuel plate which is thinner and contains more Uranium-235 than the current HEU plate. These differences, coupled with the fact that the conversion should be accomplished without serious degradation of reactor characteristics and capability, has resulted in core design studies and thermal hydraulic studies not only at the current 2 MW but also at the maximum power level of the reactor, 5 MW. In addition, during the course of its 23 years of operation, it has become clear that the main uses of the reactor are neutron scattering and neutron activation analysis. The requirement to convert to LEU presents an opportunity during the conversion to optimize the core for the utilization and to restudy the thermal hydraulics using modern techniques. This paper will present the preliminary conclusions of both aspects. (Author)

  18. A simple method for rapidly processing HEU from weapons returns

    Energy Technology Data Exchange (ETDEWEB)

    McLean, W. II; Miller, P.E.

    1994-01-01

    A method based on the use of a high temperature fluidized bed for rapidly oxidizing, homogenizing and down-blending Highly Enriched Uranium (HEU) from dismantled nuclear weapons is presented. This technology directly addresses many of the most important issues that inhibit progress in international commerce in HEU; viz., transaction verification, materials accountability, transportation and environmental safety. The equipment used to carry out the oxidation and blending is simple, inexpensive and highly portable. Mobile facilities to be used for point-of-sale blending and analysis of the product material are presented along with a phased implementation plan that addresses the conversion of HEU derived from domestic weapons and related waste streams as well as material from possible foreign sources such as South Africa or the former Soviet Union.

  19. Loading and initial start-up testing of the low-enrichment uranium core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Talnagi, J.W.

    1989-01-01

    Conversion of the Ohio State University Research Reactor (OSURR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel elements was begun in August 1985, with funding provided by the U.S. Department of Energy (DOE) and the university. Conversion of the OSURR from HEU to LEU fuel was successfully completed. The reactor is operational at 10-kW steady-state thermal power. Measurements of selected core parameters have been made and compared with predicted values and previous values for the HEU core. In general, measured results agree well with predicted performance, and minor changes have been detected in certain core parameters as a result of the change to LEU fuel. Future plans include additional core testing and a possible increase in operating power

  20. A fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1985-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  1. A fuel cycle cost study with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  2. Calculations in support of the MNR core conversion

    International Nuclear Information System (INIS)

    Day, S.E.; Butler, M.P.; Garland, Wm. J.

    2002-01-01

    Calculations and results in support of the HEU to LEU fuel conversion for the McMaster Nuclear Reactor are described. Static reactor physics studies were used to determine local and global power distributions; facilitating the definition of a Reference Core configuration for mixed HEU-LEU and complete LEU loadings. Fission product inventory calculations were used to compare the two fuel enrichments from a radiological hazard point of view. Thermalhydraulic models were created and analyzed to determine steady-state temperature distributions and safety margins, and used as a scoping tool the in development of a full core thermalhydraulic model. The behaviour of the two enrichment fuels was investigated in the context of a protected startup transient. The simulation results support the conclusion that the LEU fuel behaves in much the same way as the HEU fuel, which it is replacing. The conversion results in no new safety issues or significant changes in safety parameters. (author)

  3. Implementation of the United States/Russian HEU Agreement: Current Status and Prospects

    International Nuclear Information System (INIS)

    Rutkowski, E

    2003-01-01

    During Calendar Year (CY) 2002, the Russian Federation (R.F.) delivered low enriched uranium (LEU) from the conversion and processing of 30 metric tons (MT) of weapons-grade (90% 235 U assay) uranium. Through July 2003, the Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP) will have monitored the conversion of over 190 MT HEU into LEU. This total represents about 38 percent of the projected 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the destruction of 7,600 nuclear devices. The National Nuclear Security Administration's (NNSA) HEU-TIP monitors the processing of this HEU at four Russian uranium-processing plants. During CY 2002, United States (U.S.) personnel monitored this process for a total of 194 monitor-weeks by staffing a Transparency Monitoring Office (TMO) located in Novouralsk, and through a series of five-day Special Monitoring Visits (SMV) to the four plants. U.S. monitor observations include the inventory of in-process containers, the observation of operations and non-destructive assay measurements (NDA) to determine 235 U enrichment, as well as the examination and validation of Russian Material Control and Accountability (MC and A) documents. In addition, the U.S. designed Blend Down Monitoring System (BDMS) installed at the Ural Electrochemical Integrated Plant (UEIP) in January 1999 monitored all HEU blended at that facility, which is about 50 percent of the HEU blended into LEU during CY 2002. Recently we installed a BDMS at the Electrochemical Plant (ECP) in Zelenogorsk and plans are underway to install a BDMS at the Siberian Chemical Enterprise (SChE) in Seversk in late 2004. On a very positive note, interpersonal interactions between U.S. and Russian technical experts continues to expand and have proven to be an important element of the transparency regime. On the tenth anniversary of the HEU Purchase Agreement, the Ministry of the R.F. for Atomic Energy (Minatom) also saluted the

  4. Radiological consequence analysis with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1984-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

  5. Radiological consequence analysis with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1984-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables

  6. 31 CFR 540.305 - HEU Agreements.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false HEU Agreements. 540.305 Section 540.305 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... Federation for Atomic Energy Concerning the Transfer of Source Material to the Russian Federation signed at...

  7. A radiological consequence analysis with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1985-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and nonsite specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. (author)

  8. Overview of Russian HEU transparency issues

    International Nuclear Information System (INIS)

    Kempf, C.R.; Bieniawski, A.

    1993-01-01

    The U.S. has signed an agreement with the Russian Federation for the purchase of 500 metric tons of highly-enriched uranium (HEU) taken from dismantled nuclear weapons. The HEU will be blended down to low-enriched uranium and will be transported to the U.S. to be used by fuel fabricators to make fuel for commercial nuclear power plants. Both the U.S. and Russia have been preparing to institute transparency measures to provide assurance that nonproliferation and arms control objectives specified in the agreement are met. This paper provides background information on the original agreement and on subsequent negotiations with the Russians, as well as discussion of technical aspects of developing transparency measures suited to the facilities and processes which are expected to be involved. Transparency has been defined as those agreed-upon measures which build confidence that arms control and non-proliferation objectives shared by the parties are met. Transparency is a departure from exhaustive, detailed arms control verification regimes of past agreements, which were based on a presumption of detecting transgressions as opposed to confirming compliance

  9. IAEA Mission Sees High Commitment to Safety at Ghana's Research Reactor After HEU to LEU Fuel Conversion

    International Nuclear Information System (INIS)

    2018-01-01

    An International Atomic Energy Agency (IAEA) team of experts said the operator of Ghana’s research reactor has demonstrated a high commitment to safety following the conversion of the reactor core to use low enriched uranium (LEU) as fuel instead of high enriched uranium (HEU). The team also made recommendations for further safety enhancements. The Integrated Safety Assessment for Research Reactors (INSARR) team concluded a five-day mission today to assess the safety of the GHARR-1 research reactor, originally commissioned in 1994. The 30 kW reactor, operated by the Ghana Atomic Energy Commission (GAEC) at the National Nuclear Research Institute in the capital Accra, is used primarily for trace element analysis for industrial or agricultural purposes, research, education and training. In 2017, the reactor core was converted in a joint effort by Ghana, the United States and China, with assistance from the IAEA. The IAEA supported the operation to eliminate proliferation risks associated with HEU, while maintaining important scientific research. The team made recommendations for improvements to the GAEC, including: • Completing the revision of reactor safety and operating documents to reflect the results of the commissioning of the reactor after the core fuel conversion. • Enhancing the training and qualification programme for operating personnel. • Improving the capability for monitoring operational safety parameters under all conditions. • Strengthening radiation protection by establishing an effective radiation monitoring of workplace. The GAEC said it will request a follow-up INSARR mission by 2020.

  10. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2007-12-15

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  11. Fluxes at experiment facilities in HEU and LEU designs for the FRM-II

    International Nuclear Information System (INIS)

    Hanan, N. A.

    1998-01-01

    An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime(50 days) and the same neutron flux performance (8 x 10 14 n/cm 2 -s in the reflector). LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues that were raised by TUM have been addressed in Refs. 1-3. The conclusions of these analyses are summarized below. This paper addresses four additional issues that have been raised in several forums, including Ref 4: heat generation in the cold neutron source (CNS), the gamma and fast neutron fluxes which are components of the reactor noise in neutron scattering experiments in the experiment hall of the reactor, a fuel cycle length difference, and the reactivity worth of the beam tubes and other experiment facilities. The results show that: (a) for the same thermal neutron flux, the neutron and gamma heating in the CNS is smaller in the LEU design than in the HEU design, and cold neutron fluxes as good or better than those of the HEU design can be obtained with the LEU design; (b) the gamma and fast neutron components of the reactor noise in the experiment hall are about the same in both designs; (c) the fuel cycle length is 50 days for both designs; and (d) the absolute value of the reactivity worth of the beam tubes and other experiment facilities is smaller in the LEU design, allowing its fuel cycle length to be increased to 53 or 54 days. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical

  12. Core instrumentation and pre-operational procedures for core conversion HEU to LEU

    International Nuclear Information System (INIS)

    1984-02-01

    This report is intended for the reactor operator, to be used as a manual or checklist for general guidance on pre-startup activities that need to be addressed in preparation for conversion to Low Enriched Fuel (LEU). All nuclear, thermodynamic and safety calculations should have been performed prior to this stage of the core conversion process. During these calculations and certainly before ordering the new LEU fuel elements the reactor operator needs to very carefully consider additional important factors concerning the new fuel: fuel reliability, reliability of fuel fabricator, reprocessing contract or fuel element storage and disposal, economics of the new fuel cycle. At this stage, too, a preoperational experimental programme has to be developed and presented to the regulatory authorities for approval. This experimental programme could lead to additional requirements on: in-core instrumentation, out-of-core instrumentation or additional experimental devices. Detailed instructions on specific tests and measurements are not provided in this report since much information on the subject is available in the open literature

  13. Neutronic design of a LEU [low enriched uranium] core for the Ohio State University research reactor

    International Nuclear Information System (INIS)

    Seshadri, M.D.; Aybar, H.S.; Aldemir, T.

    1987-01-01

    The 10 kw HEU fuelled Ohio State University Reactor (OSURR) will be upgraded to operate at 500 kW with standardized 125 g 235 U LEU U 3 Si 2 fuel plates. An earlier scoping study based on two-dimensional diffusion calculations has identified the potential LEU core configurations for the conversion/upgrade of OSURR using the standardized plates in a 16-plate (+ 2 dummy plates) standard and 10-scoping study is improved for a more precise determination of the excess reactivities and safety rod worths for these potential configurations. Comparison of the results obtained by the improved model to experimental results and to the results of full-core Monte Carlo simulations shows excellent agreement. The results also indicate that the conversion/upgrade of OSURR can be realized with three possible LEU core configurations while maintaining a cold, clean shutdown margin of 1.57-1.91 % Δ k/k, depending on the configuration used. (Author)

  14. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

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

    1983-01-01

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  15. HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

    International Nuclear Information System (INIS)

    1995-09-01

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U 3 O 8 ) per year to produce 177 MT of 4% U235 assay U 3 O 8 , for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams

  16. Ion-induced gammas for photofission interrogation of HEU.

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, Barney Lee (Sandia National Laboratories, Albuquerque, NM); Antolak, Arlyn J.; Morse, Daniel H.; Provencio, Paula Polyak (Sandia National Laboratories, Albuquerque, NM)

    2006-03-01

    High-energy photons and neutrons can be used to actively interrogate for heavily shielded special nuclear material (SNM), such as HEU (highly enriched uranium), by detecting prompt and/or delayed induced fission signatures. In this work, we explore the underlying physics for a new type of photon source that generates high fluxes of mono-energetic gamma-rays from low-energy (<500 keV) proton-induced nuclear reactions. The characteristic energies (4- to 18-MeV) of the gamma-rays coincide with the peak of the photonuclear cross section. The source could be designed to produce gamma-rays of certain selected energies, thereby improving the probability of detecting shielded HEU or providing a capability to determine enrichment inside sealed containers. The fundamental physics of such an interrogation source were studied in this LDRD through scaled ion accelerator experiments and radiation transport modeling. The data were used to assess gamma and neutron yields, background, and photofission-induced signal levels from several (p,{gamma}) target materials under consideration.

  17. Department of Energy HEU ES and H vulnerability assessment, Savannah River Site, Site Assessment Team report. Revision 2

    International Nuclear Information System (INIS)

    Geddes, R.L.; Barone, A.; Shook, H.E. Varner, C.E.; Rollins, R.

    1996-01-01

    This report fulfills the directive issued by the Secretary of Energy on February 22, 1996 to complete a comprehensive assessment of potential vulnerabilities associated with the management of highly enriched uranium (HEU) throughout the DOE complex. In a subsequent letter instruction, the DOE-SR Field Office formally directed WSRC to conduct an assessment of the HEU materials at SRS. The term ''ES and H vulnerabilities'' is defined for the purpose of this assessment to mean conditions or weaknesses that could lead to unnecessary or increased exposure of workers or the public to radiation or to HEU-associated chemical hazards, or to the release of radioactive materials to the environment. The assessment will identify and prioritize ES and H vulnerabilities, and will serve as an information base for identifying corrective actions for the safe management of HEU. Primary facilities that hold HEU at SRS are H-Canyon, K-Reactor assembly area, K, L, and P-Reactor disassembly basins, and the Receiving Basin for Offsite Fuels (RBOF)

  18. HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

    International Nuclear Information System (INIS)

    1995-09-01

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

  19. HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  20. Passive Time Coincidence Measurements with HEU and DU Metal Castings

    International Nuclear Information System (INIS)

    McConchie, Seth M.; Hausladen, Paul; Mihalczo, John T.; Wright, Michael C.; Archer, Daniel E.

    2008-01-01

    A Department of Energy sponsored Oak Ridge National Laboratory/Y-12 National Security Complex program of passive time coincidence measurements has been initiated at Y-12 to evaluate the ability to determine the presence of high enriched uranium (HEU) and distinguish it from depleted uranium (DU). This program uses the Nuclear Materials Identification System (NMIS) without an active interrogation source. Previous passive NMIS measurements with Pu metal and Pu oxide have been successful in determining the Pu mass, assuming a known 240Pu content. The spontaneous fission of uranium metal is considerably lower than Pu and measurements of this type have been performed at Lawrence Livermore National Laboratory. This work presents results of measurements of HEU and DU metal castings using moderated 3He detectors.

  1. Some considerations about standardization

    Energy Technology Data Exchange (ETDEWEB)

    Dewez, Ph L; Fanjas, Y R [C.E.R.C.A., Romans (France)

    1985-07-01

    Complete standardization of research reactor fuel is not possible. However the transition from HEU to LEU should be an opportunity for a double effort towards standardization and optimization in order to reduce cost. (author)

  2. Some considerations about standardization

    International Nuclear Information System (INIS)

    Dewez, Ph.L.; Fanjas, Y.R.

    1985-01-01

    Complete standardization of research reactor fuel is not possible. However the transition from HEU to LEU should be an opportunity for a double effort towards standardization and optimization in order to reduce cost. (author)

  3. HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U{sub 3} O{sub 8}) per year to produce 177 MT of 4% U235 assay U{sub 3} O{sub 8}, for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams.

  4. ADS with HEU in the Vinca Institute

    International Nuclear Information System (INIS)

    Pesic, M.; Sobolevsky, N.

    2000-01-01

    The 'Conceptual design of ADS' is a new project proposed in the Vin.a Institute for the next three years. In this paper, an option in the project - an idea of high-enriched uranium (HEU) - H 2 O low-flux ADS is shown. Preliminary results of design study and calculations of the beam-target interaction and neutronics of proposed sub-critical system are given. (author)

  5. The whole-core LEU fuel demonstration in the ORR

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Bretscher, M.M.; Cornella, R.J.; Hobbs, R.W.

    1985-01-01

    A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worth, cycle length, fuel discharge burn-up, gamma heating rate, β eff /l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed. Key issues being addressed in the safety assessment are fuel performance, radiological consequences, margin to burnout and transient behavior. The LEU core is comparable in all safety aspects to the HEU core and the transition core is only marginally worse owing to higher power seeking factors. (author)

  6. Critical experiments of JMTRC MEU cores

    International Nuclear Information System (INIS)

    Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

    1984-01-01

    The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

  7. HEU to LEU Conversion and Blending Facility: UNH blending alternative to produce LEU UNH for commercial use

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form that is more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed. This document provides data to be used in the environmental impact analysis for the UNH blending HEU disposition option. Process requirements, resource needs, employment needs, waste/emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.

  8. HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation.

  9. HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

    International Nuclear Information System (INIS)

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation

  10. Core labour standards and exports

    OpenAIRE

    Siroën, Jean-Marc

    2017-01-01

    (english) Core labour standards defined by the ILO in 1998 are universal, but applied very differently across countries. Compliance is much higher in high income countries. However, the causality between improved labour standards and economic growth remains a controversial issue. Export-led growth strategies might encourage developing countries to curb the process of standards improvement. In this way, they can raise the volume of their unskilled labour endowments (child and/or forced labour)...

  11. Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

    As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

  12. Neutronic analysis of HEU to LEU conversion calculation for AEOI 5 MW pool-type MTR fuel research reactor core

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Lutz, D.; Bartsch, G.

    1987-07-01

    The possibility of converting HEU(93%) fuel to LEU(20%) fuel without or with slight alteration to the fuel element geometry is discussed. The fuel density varies between 1.7 to 4.1 g U-235/cm. In cross section generation a unit cell with an extra zone to account for extra Al and water was considered. In burnup calculations a sequential shuffling pattern was assumed with fixed position control fuel elements. A cross section data set in 45 energy groups were generated using RSYST/CGM system using the cross section library JFET. Then for 2D-diffusion calculations homogenized and condensed 5 energy group cross sections were prepared. (orig./HP)

  13. The conversion of NRU from HEU to LEU fuel

    International Nuclear Information System (INIS)

    Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

    1990-01-01

    The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

  14. Performance of an active well coincidence counter for HEU samples

    International Nuclear Information System (INIS)

    Ferrari, Francesca; Peerani, Paolo

    2010-01-01

    Neutron coincidence counting is the reference NDA technique used in nuclear safeguards to measure the mass of nuclear material in samples. For high-enriched uranium (HEU) samples active neutron interrogation is generally performed and the most common device used by nuclear inspectors is the Active Well Coincidence Counter (AWCC). Within her master thesis at the Polytechnic of Milan, the first author performed an intensive study on the characteristics and performances of the AWCC in order to assess the 235 U mass in HEU oxide samples at the PERLA laboratory of JRC. The work has been summarised in this paper that starts with the optimisation of the use of AWCC for nuclear safeguards, describing the calibration procedure, reporting results of a series of verification measurements, summarising the performances that can be obtained with this instruments during inspections at fuel production plants and concluding with the discussion of uncertainties related to these measurements.

  15. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  16. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  17. What the difference to use LEU and HEU fuel elements separately or together in a research reactor

    International Nuclear Information System (INIS)

    Kaya, S.; Uestuen, G.

    2005-01-01

    Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

  18. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

    1982-07-01

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  19. Darwin Core: An Evolving Community-Developed Biodiversity Data Standard

    Science.gov (United States)

    Wieczorek, John; Bloom, David; Guralnick, Robert; Blum, Stan; Döring, Markus; Giovanni, Renato; Robertson, Tim; Vieglais, David

    2012-01-01

    Biodiversity data derive from myriad sources stored in various formats on many distinct hardware and software platforms. An essential step towards understanding global patterns of biodiversity is to provide a standardized view of these heterogeneous data sources to improve interoperability. Fundamental to this advance are definitions of common terms. This paper describes the evolution and development of Darwin Core, a data standard for publishing and integrating biodiversity information. We focus on the categories of terms that define the standard, differences between simple and relational Darwin Core, how the standard has been implemented, and the community processes that are essential for maintenance and growth of the standard. We present case-study extensions of the Darwin Core into new research communities, including metagenomics and genetic resources. We close by showing how Darwin Core records are integrated to create new knowledge products documenting species distributions and changes due to environmental perturbations. PMID:22238640

  20. On the importance of ending the use of HEU in the nuclear fuel cycle: An updated assessment

    International Nuclear Information System (INIS)

    Glaser, Alexander; Hippel, Frank von

    2002-01-01

    The events of September 2001 have created a renewed urgency with regard to the disposition and future use and management of nuclear-weapons-usable materials. Highly enriched uranium (HEU) has received particular attention because it is relatively easy to use in a nuclear weapon and therefore an obvious candidate for diversion or theft by state or nonstate actors. The role of the RERTR program in this context and its contribution to global security can hardly be overemphasized. This article reviews existing or proposed activities to reduce the threat posed by HEU, how these activities are linked to the RERTR program, and outlines the most urgent steps to be taken to approach the ultimate objective of eliminating non-weapons HEU inventories in the world. (author)

  1. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab

  2. Conversion and standardization of university reactor fuels using low-enrichment uranium - Options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The U.S. Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the U.S. Department of Energy. (author)

  3. Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U3Si2 in RECH-1

    International Nuclear Information System (INIS)

    Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J.

    1999-01-01

    The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U 3 Si 2 -Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (author)

  4. Shipment of VINCA Institute's HEU fresh fuel to Russia

    International Nuclear Information System (INIS)

    Pesic, Milan; Sotic, Obrad

    2002-01-01

    This paper shows, for the first time, the basic data related to the recent shipment of the fresh HEU fuel elements from Yugoslavia back to Russia for uranium down blending. In this way, Yugoslavia gives its contribution to the RERTR program and to the world's joint efforts to prevent possible terrorist action against nuclear material potentially usable for production of nuclear weapons. (author)

  5. HEU age determination by the activity ratio {sup 227}Th/{sup 235}U

    Energy Technology Data Exchange (ETDEWEB)

    Li, Junjie; Zeng, Lina; Wu, Jian; Zheng, Chun; Li, Jiansheng, E-mail: lastljj@hotmail.com

    2014-02-15

    It is important to measure the age of a highly enriched uranium (HEU) assembly for authentication of the material in the frame of arms control inspections. A new non-destructive gamma spectrometric method for HEU age-dating is reported. This method relies on measuring the daughter/parent activity ratio {sup 227}Th/{sup 235}U by high-resolution gamma spectrometry. Only a narrow gamma range of energy of uranium from 230 keV to 242 keV will be used for analysis. The relative efficiency of every characteristic gamma ray changes in a small range because it has a near energy, which makes the results more accurate in theory. It provides a quick and reliable method for HEU age determination. Several gamma spectra of the same HEU assembly have been measured with different conditions (gain settings, distance and measurement time). When a branching ratio of 12.6% was chosen for the 235.96 keV line of {sup 227}Th, we obtained the activity ratios of (5.61 ± 0.40) × 10{sup −4}, (5.17 ± 0.39) × 10{sup −4}, (5.26 ± 0.39) × 10{sup −4}, (5.10 ± 0.35) × 10{sup −4}, (5.50 ± 0.44) × 10{sup −4} and (5.47 ± 0.42) × 10{sup −4}, respectively. These ratios correspond to ages of 52.2 ± 2.4 years, 49.7 ± 2.3 years, 50.1 ± 2.3 years, 49.3 ± 2.2 years, 51.6 ± 2.5 years and 51.5 ± 2.4 years, respectively, which are consistent with the known age of this material and the results of the U–Bi method.

  6. Ohio State University Nuclear Reactor Laboratory HEU fuel shipment summary. Final

    International Nuclear Information System (INIS)

    1997-01-01

    In November 1988, OSURR converted from HEU fuel to LEU fuel. As a result they needed to get rid of their HEU fuel by shipping it to Savannah River. The players in the fuel shipping game are: OSURR as the keeper of the fuel; DOE as the owner of fuel and shipper of record; Tri-State Motor Transit Co. for transporting the cask; Muth Brothers as the rigger responsible for getting the cask on and off the truck and in and out of the building; Hoffman LaRoche/Cintichem as the owner of the cask; Savannah River as the receiver of the fuel; and the NRC for approval of the Security Plan, QA Plan, etc. This report gives a chronological history of the events from February 1989 to June 1, 1995, the actual day of shipment. The cask was received at Savannah River on June 2, 1995

  7. Advocacy: Emphasizing the Uncommon about the Common Core State Standards

    Science.gov (United States)

    Kaplan, Sandra N.

    2014-01-01

    The author describes key issues and uncommon concerns about the Common Core State Standards that fit within two categories: philosophical and pedagogical. Philosophically, Common Core State K-12 Standards should not be expected to be mastered at a specific grade level but based on developmental readiness. Pedagogically, Common Core State Standards…

  8. Orsphere: Physics Measurments For Bare, HEU(93.2)-Metal Sphere

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); White, Christine E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dyrda, James P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tancock, Nigel P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files” (Reference 1). While performing the ORSphere experiments care was taken to accurately document component dimensions (±0.0001 inches), masses (±0.01 g), and material data. The experiment was also set up to minimize the amount of structural material in the sphere proximity. Two, correlated spheres were evaluated and judged to be acceptable as criticality benchmark experiments. This evaluation is given in HEU-MET-FAST-100. The second, smaller sphere was used for additional reactor physics measurements. Worth measurements (Reference 1, 2, 3 and 4), the delayed neutron fraction (Reference 3, 4 and 5) and surface material worth coefficient (Reference 1 and 2) are all measured and judged to be acceptable as benchmark data. The prompt neutron decay (Reference 6), relative fission density (Reference 7) and relative neutron importance (Reference 7) were measured, but are not evaluated. Information for the evaluation was compiled from References 1 through 7, the experimental logbooks 8 and 9 ; additional drawings and notes provided by the experimenter; and communication with the lead experimenter, John T. Mihalczo.

  9. Voter Perceptions: Common Core State Standards & Tests

    Science.gov (United States)

    Achieve, Inc., 2014

    2014-01-01

    Since June 2010, 46 states and Washington DC have adopted the Common Core State Standards (CCSS)--K-12 standards in mathematics and English language arts/literacy developed through a multi-state initiative led by the National Governors Association and the Council of Chief State School Officers. Implementation of the standards is underway in all of…

  10. Core management and reactor physics aspects of the conversion of the NRU reactor to LEU

    International Nuclear Information System (INIS)

    Atfield, M.D.

    1985-01-01

    Results of work done to assess the effects of converting the NRU reactor to LEU are presented. The effects are small, and the operational rules and safety analysis, appropriate to the HEU core, will still apply. (author)

  11. HEU to LEU Conversion and Blending Facility: UF6 blending alternative to produce LEU UF6 for commercial use

    International Nuclear Information System (INIS)

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF 6 to produce a UF 6 product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF 6 blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed

  12. Common Core Science Standards: Implications for Students with Learning Disabilities

    Science.gov (United States)

    Scruggs, Thomas E.; Brigham, Frederick J.; Mastropieri, Margo A.

    2013-01-01

    The Common Core Science Standards represent a new effort to increase science learning for all students. These standards include a focus on English and language arts aspects of science learning, and three dimensions of science standards, including practices of science, crosscutting concepts of science, and disciplinary core ideas in the various…

  13. Moderator temperature effects on reactivity of HEU core of MNSR

    International Nuclear Information System (INIS)

    Ahmad, Siraj-ul-Islam; Sahibzada, Tasveer Muhammad

    2012-01-01

    Highlights: ► The MNSR core was analyzed to see the cross section effects on moderator temperature coefficient of reactivity. ► WIMS-D code was used for cell calculations. ► The 3D diffusion theory code PRIDE was first validated using IAEA benchmark problem and then used for analysis of MNSR. ► The differences among results for various libraries were discussed. -- Abstract: In this article we report on analyses that were performed to investigate the influence of cross section differences among libraries released by various centers on reactivity of Miniature Neutron Source Reactors. The 3D model of the core was developed with WIMS-D and PRIDE codes and six cross section libraries were used including JENDL-3.2, JEF-2.2, JEFF-3.3, ENDF/B-VI and ENDF/B-VII, and IAEA library. It was observed that all the libraries predict the reactivity within 10%, with IAEA library giving minimum reactivity worth, and JEF-2.2 data library resulted in highest worth.

  14. Using Digital Video Production to Meet the Common Core Standards

    Science.gov (United States)

    Nichols, Maura

    2012-01-01

    The implementation of the Common Core Standards has just begun and these standards will impact a generation that communicates with technology more than anything else. Texting, cell phones, Facebook, YouTube, Skype, etc. are the ways they speak with their friends and the world. The Common Core Standards recognize this. According to the Common Core…

  15. Highly enriched uranium (HEU) politics: An enigma wrapped up in a warhead and boxed in political chaos

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    It could be fairly said that while the Cold War arose in an atmosphere of mutual mistrust and hostility, it is ending with an equal amount of confusion and uncertainty. More than a year has passed since the US and Russia signed a tentative HEU agreement in August 1992. Many of the details have been worked out, but major questions remain. And they're not just on the Russian side. The fine points of President Clinton's overall nuclear policy are only now beginning to emerge. In his first major foreign policy address, before the United Nations in late September, Clinton called for a worldwide ban on the production of plutonium and HEU for nuclear weapons. open-quotes Growing global stockpiles of plutonium and highly enriched uranium are raising the danger of nuclear terrorism for all nations,close quotes said Clinton before the UN. open-quotes We will press for an international agreement that would ban production of these materials for weapons forever.close quotes As the veil lifts from Clinton's nuclear policy, it appears the Administration realizes that Russia may have more HEU than originally thought. That possibility has been confirmed by Minatom Minister Mikhailov's disclosures to the NUKEM Market Report, which brought a greater degree of certainty to estimates that had been floating around for some time. When the Bush Administration signed the HEU pact, it apparently thought the 500 metric tons comprised most of the former Soviet Union's nuclear arsenal. Now that the number appears higher, Clinton may propose to accelerate and enlarge the HEU deal. He is due to summit with Yeltsin, if Yeltsin survives, next spring. The 500-metric-ton deal may only be the first step

  16. A comparison of the radiological consequences of a HEU and LEU fueled research reactor

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1985-01-01

    An analysis of the design basis accident radiological consequences of the HEU and LEU fueled Greek Research Reactor is presented. Doses and individual cancer risk from exposure to the passing radioactive cloud are estimated up to a distance of 20 km from the reactor site. Collective exposure and latent health effects are estimated for the total Athens area of 3081000 inhabitants. The results indicate that the plutonium isotopes buildup in the LEU fuel does not increase appreciably the consequences in respect to the HEU fueled reactor. The plutonium impact concerns mainly bone effects and secondly lung and whole body effects. The contribution to the limiting thyroid dose and the corresponding thyroid effects is insignificant. (author)

  17. Operation of automated NDA instruments for in-line HEU accounting at Y-12

    International Nuclear Information System (INIS)

    Russo, P.A.; Strittmatter, R.B.; Sandford, E.L.; Jeter, I.W.; McCullough, E.; Bowers, G.L.

    1983-01-01

    Two automated nondestructive assay instruments developed at Los Alamos in support of nuclear materials accounting needs are currently operating in-line at the Y-12 Plant for recovery of highly enriched uranium. One instrument provides the HEU inventory in the secondary solvent extraction system, and the other monitors HEU concentration in the secondary intermediate evaporator. Both instruments were installed in December 1982. Operational evaluation of these instruments has been a joint effort of Y-12 and Los Alamos. This has included comparison of the solvent extraction system inventories with direct measurement performed on the dumped solution components of the solvent extraction system, as well as comparisons of concentration assay results with the external assays of samples withdrawn from the process. The function, design, and preliminary results of the operational evaluation are reported

  18. ORSPHERE: CRITICAL, BARE, HEU(93.2)-METAL SPHERE

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-09-01

    In the early 1970’s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950’s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files.” (Reference 1) While performing the ORSphere experiments care was taken to accurately document component dimensions (±0. 0001 in. for non-spherical parts), masses (±0.01 g), and material data The experiment was also set up to minimize the amount of structural material in the sphere proximity. A three part sphere was initially assembled with an average radius of 3.4665 in. and was then machined down to an average radius of 3.4420 in. (3.4425 in. nominal). These two spherical configurations were evaluated and judged to be acceptable benchmark experiments; however, the two experiments are highly correlated.

  19. Multilateral nonproliferation cooperation: US - Led effort to remove HEU/LEU fresh and spent fuels from the Republic of Georgia to Dounreay, Scotland

    International Nuclear Information System (INIS)

    Shelton, Thomas A.; Viebrock, James M.; Riedy, Alexander W.; Moses, Stanley D.; Bird, Helen M.

    1998-01-01

    This paper presents the efforts led by United States for removing HEU/LEU fresh and spent fuel from dhe Republic of Georgia to Dounreay, Scotland. These efforts are resulted from a plan approved by the United States Government, in cooperation with the United Kingdom and Georgia Governments to rapidly retrieve and transport circa 4.3 kilograms of enriched uranium. This material consisted largely of highly enriched uranium (HEU) and a small amount of low enriched uranium (LEU) fresh fuel, as well as about 800 grams of HEU/LEU-based spent fuel from a shutdown IR T-M research reactor on the outskirts of Table's, Georgia. The technical team lead by DOE consisted of HEU handling, packaging and transportation experts from the Oak Ridge Y-12 plant, managed and operated by Lockheed Martin Energy Systems, and fuel handling and transportation experts from Nac International in Norcross, Georgia, United States

  20. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  1. Modeling in the Common Core State Standards

    Science.gov (United States)

    Tam, Kai Chung

    2011-01-01

    The inclusion of modeling and applications into the mathematics curriculum has proven to be a challenging task over the last fifty years. The Common Core State Standards (CCSS) has made mathematical modeling both one of its Standards for Mathematical Practice and one of its Conceptual Categories. This article discusses the need for mathematical…

  2. Core physics analysis in support of the FNR HEU-LEU demonstration experiment

    Energy Technology Data Exchange (ETDEWEB)

    Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

  3. Core physics analysis in support of the FNR HEU-LEU demonstration experiment

    International Nuclear Information System (INIS)

    Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

    1983-01-01

    A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

  4. Coordination Between the HEU Transparency Program and the Material Protection, Control and Accountability Program

    International Nuclear Information System (INIS)

    Glaser, J.; Hernandez, J.; Dougherty, D.; Bieniawski, A.; Cahalane, P.; Mastal, E.

    2000-01-01

    DOE sponsored programs such as Material Protection Control and Accountability (MPC and A) and implementation of the Highly-Enriched Uranium (HEU) Transparency Program send US personnel into Russian nuclear facilities and receive Russian representatives from these programs. While there is overlap in the Russian nuclear facilities visited by these two programs, there had not been any formal mechanism to share information between them. Recently, an MPC and A/HEU Working Group was developed to facilitate the sharing of appropriate information and to address concerns expressed by Minatom and Russian facility personnel such as US visit scheduling conflicts. This paper discusses the goals of the Working Group and ways it has helped to allow the programs to work more efficiently with the Russian facilities

  5. 76 FR 54537 - Swap Data Repositories: Registration Standards, Duties and Core Principles

    Science.gov (United States)

    2011-09-01

    ... Part 49 Swap Data Repositories: Registration Standards, Duties and Core Principles; Final Rule #0;#0...: Registration Standards, Duties and Core Principles AGENCY: Commodity Futures Trading Commission. ACTION: Final... registration requirements, statutory duties, core principles and certain compliance obligations for registered...

  6. Slicing and Dicing the ELA Common Core Standards

    Science.gov (United States)

    Goatley, Virginia

    2012-01-01

    The English Language Arts Common Core State Standards (ELA CCSS) come at a time when many reading teachers, literacy coaches, and classroom teachers seek more extensive literacy practices than the policy mandates of No Child Left Behind and Reading First. These initiatives placed requirements for instruction in core aspects of reading at the…

  7. HEU to LEU Conversion and Blending Facility: UF{sub 6} blending alternative to produce LEU UF{sub 6} for commercial use

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF{sub 6} to produce a UF{sub 6} product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF{sub 6} blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.

  8. Thermal hydraulic model validation for HOR mixed core fuel management

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

    1997-01-01

    A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

  9. Forensic analysis of a smuggled HEU sample interdicted in Bulgaria

    International Nuclear Information System (INIS)

    Niemeyer, S.; Hutcheon, I.

    2002-01-01

    Full text: A sample of HEU was seized in Rousse, Bulgaria on May 29, 1999, at a border crossing between Bulgaria and Romania. A search of the suspect's vehicle uncovered a lead canister hidden in the car trunk. The initial examination of the contents by Bulgarian scientists indicated that the sample was indeed HEU, and subsequently arrangements were made for a U.S. team of nuclear forensics scientists from several national laboratories to conduct a thorough examination. This report gives a summary of the results. The HEU sample was contained in a glass ampoule that was embedded in a yellow wax, and in turn the wax filled the inside of the cylindrical lead container. A broad set of techniques was used to examine both nuclear and non-nuclear materials. Our general experimental approach has been previously described at meetings of the Nuclear Smuggling International Technical Working Group (ITWG), but this case represents the application of the most diverse set of nuclear forensic measurements for an actual seized sample. Analysis of the HEU itself included particle characterization, stoichiometry, impurity elements, residual nuclides, age-dating, and U and Pu isotopics. Measurements by XRD, SEM, and TEM show that sample is mostly U3O8, with minor amounts of two other phases. The powder is extremely fine-grained (160 nm mean) and quite uniform in size. Most grains (95%) are equidimensional, with the remainder rod-or plate-shaped. The U is 72.7% U-235 with a high U-236 abundance of 12.1%. The sample is reprocessed, reactor-irradiated material. The original U enrichment was 90% and the irradiation burned up about 50% of the initial U-235. Pu is present at a very low-level (3 ppb); the Pu-239 abundance is 82% with 240/239=0.12. Three fission products were detected at low levels, giving unambiguous evidence of fuel recycling. The total impurity content is about 600 ppmw (mostly S, Cl, Fe, and Br), which we interpret as indicating a batch processing operation because the

  10. Potential Ramifications of Common Core State Standards Adoption on Information Literacy

    Directory of Open Access Journals (Sweden)

    Jacob Paul Eubanks

    2014-07-01

    Full Text Available In the United States, the decline in jobs for high school educated workers and the proliferation of jobs for post-secondary educated workers is driving the development of the Common Core State Standards. The Common Core State Standards theoretically shift K-12 pedagogy towards ability development of critical and extended thinking skills, preparing high school graduates for college and career readiness. This literature review explores the reasoning behind the shift to the Common Core State Standards and asks questions regarding the potential ramifications their adoption might have on post-secondary information literacy instruction.

  11. Making It Happen: Common Core Standards

    Science.gov (United States)

    National Council of Teachers of Mathematics, 2011

    2011-01-01

    This one-of-a-kind guide identifies and highlights the ways in which NCTM (National Council of Teachers of Mathematics) resources can support teachers as they implement and supplement the Common Core State Standards for Mathematics (CCSSM) in their states. The guide and accompanying charts are tools to help educators as they continue to make…

  12. Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1995-01-01

    This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

  13. Licensing procedures and safety criteria for core conversion in Japan

    International Nuclear Information System (INIS)

    Kanda, K.; Nakagome, Y.; Hayashi, M.

    1983-01-01

    In Japan, the establishment and operation of nuclear installations are governed mainly by the Law for Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors. This law lays down the regulations and conditions for licensing of the various installations involved in the nuclear fuel cycle, namely licensing of installations for refining, fabricating and reprocessing; and reactors, as well as licensing of the use of nuclear fuels in research facilities. Although procedures for the installations listed above vary depending on the installation concerned, only those relating to construction and operation of reactor facilities will be analysed in this study, as the conditions and principles applying to licensing and control of other installations are, to a large extent, similar to those concerning reactor facilities. The second part of this presentation describes the safety review of the KUCA reactor core conversion form HEU to MEU. For the safety review of the core conversion, the Committee on Examination of Reactor Safety of Japanese Government examined mainly the the nuclear characteristics and the integrity of aluminide fuel plates, which was very severe because we had no experience to use aluminide fuel plates in Japan. The integrity of fuel plates and the results of the worst accident analysis for the MEU core are shown with the comparison between the HEU and MEU cores. The significant difference was not observed between them. All the regulatory procedures were completed in September 1980. Fabrication of MEU fuel elements for the KUCA experiments by CERCA in France was started in September 1980, and will be completed in March 1981. The critical experiments in the KUCA with MEU fuel will be started on a single-core in May 1981 as a first step. Those on a coupled-core will follow

  14. ORALLOY (93.15 235U) METAL ANNULI WITH BERYLLIUM CORE

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland M.; Reed, Raymond L.; Mihalczo, John T.

    2010-01-01

    A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, two were performed that consisted of uranium metal annuli with a solid beryllium metal core. The outer diameter of the annuli was approximately 13 or 15 inches with an inner diameter of 7 inches. The diameter of the core was approximately 7 inches. The critical height of the configurations was approximately 5 and 4 inches, respectively. The uranium annuli consisted of multiple stacked rings with diametral thicknesses of approximately 2 inches apiece and varying heights. The 15-inch experiment was performed on June 4, 1963, and the 13-inch experiment on July 12, 1963 by J. T. Mihalczo and R. G. Taylor (Ref. 1) with accompanying logbook. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast-spectra experiments were determined to represent acceptable benchmarks. The calculated eigenvalues for both the detailed and simple models are within approximately 0.6% of the benchmark values, but significantly greater than 3s from the benchmark value because the uncertainty in the benchmark is very small: eff of ∼0.67%. Unreflected and unmoderated experiments with the same highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in HEU MET

  15. Translating the Common Core State Standards

    Science.gov (United States)

    Tienken, Christopher H.; Orlich, Donald C.

    2013-01-01

    As the authors describe in Chapter 7 of their new book, "The School Reform Landscape: Fraud, Myth, and Lies," the Common Core State Standards (CCSS) initiative continues to ramble on, without evidence to support its efficacy. That is because education reform in the United States is being driven largely by ideology, rhetoric, and dogma instead of…

  16. Slope across the Curriculum: Principles and Standards for School Mathematics and Common Core State Standards

    Science.gov (United States)

    Nagle, Courtney; Moore-Russo, Deborah

    2014-01-01

    This article provides an initial comparison of the Principles and Standards for School Mathematics and the Common Core State Standards for Mathematics by examining the fundamental notion of slope. Each set of standards is analyzed using eleven previously identified conceptualizations of slope. Both sets of standards emphasize Functional Property,…

  17. Knowledge Transfer and Culture Exchange between HEU and TAMU through a Summer School on Nuclear Power Engineering

    International Nuclear Information System (INIS)

    Gao, P.; Zhang, Z.; Kurwitz, R. C.; Shao, L.

    2016-01-01

    Full text: Since 2012, Harbin Engineering University (HEU) and Texas A&M University (TAMU) hold an annual Summer School on Nuclear Engineering. By now, the activity has been held four times. Each year, 15–20 students are selected from their respective institutions and paired with a counterpart to form partners. They study lectures in the first week at HEU and tour three nuclear power plants (NPP) in the second week, visit the NPP simulators, and learn the nuclear safety culture. This activity expands the scale of international nuclear educational exchange, provide a platform for the students from different countries to communicate and exchange insights into their respective culture. (author

  18. Experimental evaluation of new NEU cores in the UVAR

    Energy Technology Data Exchange (ETDEWEB)

    Farrar, P.; Hosticka, B.; Krause, D.; Mulder, R.; Rydin, R. [Univ. of Virginia, Charlottesville, VA (United States)

    1997-08-01

    The University of Virginia began working on converting the UVAR reactor to LEU fuel in the Spring of 1986. The Safety Analysis Report was completed and submitted to the NRC in late 1989. After review, the DOE order to manufacture LEU fuel was placed at B&W in March 1992, and the new fuel was received in January 1994. The 4-by-4 fully-graphite-reflected LEU-1 core went critical on April 20, 1994, and the 4-by-5 partially-graphite-reflected operational LEU-2 core went critical on April 29. Full power was achieved on May 12, 1994. Both cores behaved very much as originally predicted. All of the old HEU fuel has been shipped to Savannah River.

  19. Disposition of Highly Enriched Uranium (HEU) and Pu from nuclear weapons

    International Nuclear Information System (INIS)

    Neff, T.L.

    1992-01-01

    Last year, as the Soviet Union began to crumble and the Bush-Gorbachev talks had advanced to consider the dismantling of actual warheads instead of mere delivery systems, Dr. Thomas L. Neff, a senior member of MIT's Center for International Studies, tinkered with the notion that one day soon the US could be buying Russian uranium from scrapped nuclear arms. He also considered the costly business of dismantlement and set to work on an ingenious proposal. The basic idea was simply to purchase the HEU from Russia using funds derived from savings in the US Department of Energy's enrichment enterprise. The proposal, now part of an umbrella agreement between the US and Russia announced in early September, promises large strategic benefits for the US in terms of both financing dismantlement (to the tune of $100 million annually), as well as political oversight for the operation itself. In the words of Dr. Neff, who made briefings to both governments on the proposal, open-quotes It's a budget-neutral, win-win solution.close quotes What follows is an illustrated, step-by-step analysis of the proposal, as well as a reprint of Dr. Neff's paper, Disposition of HEU and Pu from Nuclear Weapons, as presented to the Uranium Institute's annual symposium last month

  20. Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders

    International Nuclear Information System (INIS)

    Freeman, Corey R.; Geist, William H.

    2010-01-01

    Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF 6 spins at high velocities in centrifuges to separate the molecules containing 238 U from those containing the lighter 235 U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF 6 gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

  1. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  2. Conversion of the Worcester Polytechnic Institute nuclear reactor to low enriched uranium

    International Nuclear Information System (INIS)

    Newton, T.H. Jr.

    1991-01-01

    The Training Reactor was converted to Low-Enriched Uranium (LEU) aluminide fuel in 1988 and 1989. Tests on the Highly-Enriched Uranium (HEU) core and LEU cores were performed and comparisons made. The testing consisted of critical loading, thermal neutron flux distribution, excess reactivity, regulating blade reactivity worth, and temperature coefficient of reactivity measurement. Comparisons between the LEU and HEU showed that the critical loading configurations were somewhat different with the HEU core consisting of 24 elements and the LEU core consisting of 21 1/3 elements with excess reactivities of 0.24% ΔK/K for the HEU and 0.16% for the LEU. Thermal neutron flux distributions showed similar trends in both the LEU and HEU cores. The regulating blade worth showed a larger LEU value due to thermal peaking in the blade region and temperature coefficients showed a more negative LEU value due to Doppler broadening. Low induced activity of the HEU fuel permitted shipment to the Westinghouse Savannah River Facility using DOT-6M type B containers on 8 August, 1989. (orig.)

  3. A Creative Approach to the Common Core Standards: The Da Vinci Curriculum

    Science.gov (United States)

    Chaucer, Harry

    2012-01-01

    "A Creative Approach to the Common Core Standards: The Da Vinci Curriculum" challenges educators to design programs that boldly embrace the Common Core State Standards by imaginatively drawing from the genius of great men and women such as Leonardo da Vinci. A central figure in the High Renaissance, Leonardo made extraordinary contributions as a…

  4. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  5. UBERA-6 project: Achievements of 4 working years

    International Nuclear Information System (INIS)

    Blaumann, H; Fernandez, C; Dell Occhio, L; D Ovidio, C; Fabro, J; Miceli, M; Novara O; Perez, A; Taboada, H

    2009-01-01

    On May 2005 the President of CNEA created the UBERA-6 project, belonging to the former Technology and Environmental Management, with the aim to convert to Low Enriched Uranium (LEU) the RA-6 reactor core, to swap with the US Department of Energy (US-DoE) equivalent inventories of High Enriched Uranium (HEU) for LEU, to export to USA the spent HEU core and to recover and downblend to LEU remnant HEU inventories contained in fuel and irradiation target scraps. By means of two contracts signed by CNEA and US-DoE, acquisition of consumables and graphite reflectors, the fabrication of LEU core replacement, conditioning, transport and exportation of spent HEU core and subsequent supply of fresh LEU for fuel and irradiation targets used in our research reactors were costed. During July, 2006 468 HEU based fresh plates were exported to USA. On June 30th, 2007 the RA-6 reactor temporarily stopped working and its personnel remover the HEU core to the auxiliary pool. On November 7th the former spent HEU based core was exported to USA. During May and July, 2008 the new RA-6 reactor LEU based core and control assemblies were provided. During March, 2009 the RA-6 reactor became critical. For recovering and blending down of remnant HEU inventories, the Triple Height Laboratory (LTA) was refurbished. A Supplemental Agreement to one of the original contract between CNEA and US-DoE will financially support the refurbishment of the Radiochemical Facility Laboratory (LFR) and so reprocess irradiated HEU retained in radioisotope production filters to downblend into LEU, as well as the separation of the pair Sr90-Y90 and of Cs137 inventories for further application in Nuclear Medicine. [es

  6. Implications of Common Core State Standards on the Social Studies

    Science.gov (United States)

    Kenna, Joshua L.; Russell, William B., III.

    2014-01-01

    Social studies teachers have often been on the outside looking in during much of the era billed as the standards-based educational reform (SBER), but with the adoption and implementation of the Common Core State Standards (CCSS), social studies teachers seem to have been invited back inside. Yet, how will the standards impact social studies…

  7. Time-correlated neutron analysis of a multiplying HEU source

    Energy Technology Data Exchange (ETDEWEB)

    Miller, E.C., E-mail: Eric.Miller@jhuapl.edu [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States); Kalter, J.M.; Lavelle, C.M. [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States); Watson, S.M.; Kinlaw, M.T.; Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID (United States); Noonan, W.A. [Johns Hopkins University Applied Physics Laboratory, Laurel, MD (United States)

    2015-06-01

    The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated {sup 3}He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations.

  8. Time-correlated neutron analysis of a multiplying HEU source

    Science.gov (United States)

    Miller, E. C.; Kalter, J. M.; Lavelle, C. M.; Watson, S. M.; Kinlaw, M. T.; Chichester, D. L.; Noonan, W. A.

    2015-06-01

    The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated 3He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations.

  9. Time-correlated neutron analysis of a multiplying HEU source

    International Nuclear Information System (INIS)

    Miller, E.C.; Kalter, J.M.; Lavelle, C.M.; Watson, S.M.; Kinlaw, M.T.; Chichester, D.L.; Noonan, W.A.

    2015-01-01

    The ability to quickly identify and characterize special nuclear material remains a national security challenge. In counter-proliferation applications, identifying the neutron multiplication of a sample can be a good indication of the level of threat. Currently neutron multiplicity measurements are performed with moderated 3 He proportional counters. These systems rely on the detection of thermalized neutrons, a process which obscures both energy and time information from the source. Fast neutron detectors, such as liquid scintillators, have the ability to detect events on nanosecond time scales, providing more information on the temporal structure of the arriving signal, and provide an alternative method for extracting information from the source. To explore this possibility, a series of measurements were performed on the Idaho National Laboratory's MARVEL assembly, a configurable HEU source. The source assembly was measured in a variety of different HEU configurations and with different reflectors, covering a range of neutron multiplications from 2 to 8. The data was collected with liquid scintillator detectors and digitized for offline analysis. A gap based approach for identifying the bursts of detected neutrons associated with the same fission chain was used. Using this approach, we are able to study various statistical properties of individual fission chains. One of these properties is the distribution of neutron arrival times within a given burst. We have observed two interesting empirical trends. First, this distribution exhibits a weak, but definite, dependence on source multiplication. Second, there are distinctive differences in the distribution depending on the presence and type of reflector. Both of these phenomena might prove to be useful when assessing an unknown source. The physical origins of these phenomena can be illuminated with help of MCNPX-PoliMi simulations

  10. Preliminary investigations of a mixed standard-flip core for a TRIGA Mark II

    International Nuclear Information System (INIS)

    Ringle, John C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    Several years ago it became apparent that due to our rapidly- increasing use rate, we would need a substantial amount of new fuel by late 1974 or early 1975. After investigations and discussions with GA, we decided that FLIP fuel would best meet our requirements for maximum fuel economy and high peak pulsing power. A proposal was submitted to the AEC for fuel assistance, and late in 1973 we were awarded a grant of $61,875. This will allow us to buy 3 FLIP-fueled-follower control rods, 1 instrumented FLIP fuel element, and 26 standard FLIP elements, giving us then a mixed core of approximately one-third FLIP and two-thirds standard elements. License amendments to accommodate this change are rather straightforward; modifications to the Technical Specifications will be somewhat more involved. The largest revisions which we envision are to our Safety Analysis Report. Although a few reactors have operated with a full FLIP core, and a few others have converted to mixed standard-FLIP cores, none of these has a standard Mark II core configuration. Those who have already converted to a mixed core have data and calculations which may be helpful to us, but the extent to which we can use these remains to be seen. The present status of our investigations into the analysis of a mixed standard-FLIP core will be presented. Any problems in calculational methods, finding appropriate data, modifications to Technical Specifications, etc., will be identified, and suggestions and help in these areas will be welcomed. (author)

  11. Preliminary investigations of a mixed standard-flip core for a TRIGA Mark II

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, John C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    Several years ago it became apparent that due to our rapidly- increasing use rate, we would need a substantial amount of new fuel by late 1974 or early 1975. After investigations and discussions with GA, we decided that FLIP fuel would best meet our requirements for maximum fuel economy and high peak pulsing power. A proposal was submitted to the AEC for fuel assistance, and late in 1973 we were awarded a grant of $61,875. This will allow us to buy 3 FLIP-fueled-follower control rods, 1 instrumented FLIP fuel element, and 26 standard FLIP elements, giving us then a mixed core of approximately one-third FLIP and two-thirds standard elements. License amendments to accommodate this change are rather straightforward; modifications to the Technical Specifications will be somewhat more involved. The largest revisions which we envision are to our Safety Analysis Report. Although a few reactors have operated with a full FLIP core, and a few others have converted to mixed standard-FLIP cores, none of these has a standard Mark II core configuration. Those who have already converted to a mixed core have data and calculations which may be helpful to us, but the extent to which we can use these remains to be seen. The present status of our investigations into the analysis of a mixed standard-FLIP core will be presented. Any problems in calculational methods, finding appropriate data, modifications to Technical Specifications, etc., will be identified, and suggestions and help in these areas will be welcomed. (author)

  12. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) 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) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  13. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1984-01-01

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  14. Core Knowledge and Standards: A Conversation with E.D. Hirsch, Jr.

    Science.gov (United States)

    O'Neil, John

    1999-01-01

    Hirsch believes it is vitally important to specify the "core knowledge" that all students must learn. Here, Hirsch explains elements of his K-8 core-knowledge sequence. Teachers should avoid canned lessons but should know where they are going. New English standards are unacceptable, since they omit Shakespeare's works. (MLH)

  15. Training in Vocational Assessment: Preparing Rehabilitation Counselors and Meeting the Requirements of the CORE Standards

    Science.gov (United States)

    Tansey, Timothy N.

    2008-01-01

    Assessment represents a foundational component of rehabilitation counseling services. The revised Council on Rehabilitation Education (CORE) standards implemented in 2004 resulted in the redesign of the knowledge and outcomes under the Assessment standard. The author reviews the current CORE standard for training in assessment within the context…

  16. Core Outcome Set-STAndards for Development: The COS-STAD recommendations.

    Directory of Open Access Journals (Sweden)

    Jamie J Kirkham

    2017-11-01

    Full Text Available The use of core outcome sets (COS ensures that researchers measure and report those outcomes that are most likely to be relevant to users of their research. Several hundred COS projects have been systematically identified to date, but there has been no formal quality assessment of these studies. The Core Outcome Set-STAndards for Development (COS-STAD project aimed to identify minimum standards for the design of a COS study agreed upon by an international group, while other specific guidance exists for the final reporting of COS development studies (Core Outcome Set-STAndards for Reporting [COS-STAR].An international group of experienced COS developers, methodologists, journal editors, potential users of COS (clinical trialists, systematic reviewers, and clinical guideline developers, and patient representatives produced the COS-STAD recommendations to help improve the quality of COS development and support the assessment of whether a COS had been developed using a reasonable approach. An open survey of experts generated an initial list of items, which was refined by a 2-round Delphi survey involving nearly 250 participants representing key stakeholder groups. Participants assigned importance ratings for each item using a 1-9 scale. Consensus that an item should be included in the set of minimum standards was defined as at least 70% of the voting participants from each stakeholder group providing a score between 7 and 9. The Delphi survey was followed by a consensus discussion with the study management group representing multiple stakeholder groups. COS-STAD contains 11 minimum standards that are the minimum design recommendations for all COS development projects. The recommendations focus on 3 key domains: the scope, the stakeholders, and the consensus process.The COS-STAD project has established 11 minimum standards to be followed by COS developers when planning their projects and by users when deciding whether a COS has been developed using

  17. Core Outcome Set-STAndards for Development: The COS-STAD recommendations.

    Science.gov (United States)

    Kirkham, Jamie J; Davis, Katherine; Altman, Douglas G; Blazeby, Jane M; Clarke, Mike; Tunis, Sean; Williamson, Paula R

    2017-11-01

    The use of core outcome sets (COS) ensures that researchers measure and report those outcomes that are most likely to be relevant to users of their research. Several hundred COS projects have been systematically identified to date, but there has been no formal quality assessment of these studies. The Core Outcome Set-STAndards for Development (COS-STAD) project aimed to identify minimum standards for the design of a COS study agreed upon by an international group, while other specific guidance exists for the final reporting of COS development studies (Core Outcome Set-STAndards for Reporting [COS-STAR]). An international group of experienced COS developers, methodologists, journal editors, potential users of COS (clinical trialists, systematic reviewers, and clinical guideline developers), and patient representatives produced the COS-STAD recommendations to help improve the quality of COS development and support the assessment of whether a COS had been developed using a reasonable approach. An open survey of experts generated an initial list of items, which was refined by a 2-round Delphi survey involving nearly 250 participants representing key stakeholder groups. Participants assigned importance ratings for each item using a 1-9 scale. Consensus that an item should be included in the set of minimum standards was defined as at least 70% of the voting participants from each stakeholder group providing a score between 7 and 9. The Delphi survey was followed by a consensus discussion with the study management group representing multiple stakeholder groups. COS-STAD contains 11 minimum standards that are the minimum design recommendations for all COS development projects. The recommendations focus on 3 key domains: the scope, the stakeholders, and the consensus process. The COS-STAD project has established 11 minimum standards to be followed by COS developers when planning their projects and by users when deciding whether a COS has been developed using reasonable

  18. Tying Together the Common Core of Standards, Instruction, and Assessments

    Science.gov (United States)

    Phillips, Vicki; Wong, Carina

    2010-01-01

    Clear, high standards will enable us to develop an education system that ensures that high school graduates are ready for college. The Bill & Melinda Gates Foundation has been working with other organizations to develop a Common Core of Standards. The partners working with the foundation are developing tools that will show teachers what is…

  19. A Case for Common Core State Standards: Gifted Curriculum 3.0

    Science.gov (United States)

    VanTassel-Baska, Joyce

    2012-01-01

    The Common Core State Standards (CCSS) is the most successful attempt to gain consensus across states for 21st century standards in language arts and mathematics. So far, 46 states have accepted these standards, with two consortia organized to translate them into resources and sample activities. A consultant firm has been hired to develop the…

  20. Whole-core LEU fuel demonstration in the ORR

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Bretscher, M.M.; Cornella, R.J.; Hobbs, R.W.

    1985-01-01

    A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worths, cycle length, fuel discharge burnup, gamma heating rates, β/sub eff/l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed

  1. The effect of core configuration on temperature coefficient of reactivity in IRR-1

    Energy Technology Data Exchange (ETDEWEB)

    Bettan, M.; Silverman, I.; Shapira, M.; Nagler, A. [Soreq Nuclear Research Center, Yavne (Israel)

    1997-08-01

    Experiments designed to measure the effect of coolant moderator temperature on core reactivity in an HEU swimming pool type reactor were performed. The moderator temperature coefficient of reactivity ({alpha}{sub {omega}}) was obtained and found to be different in two core loadings. The measured {alpha}{sub {omega}} of one core loading was {minus}13 pcm/{degrees}C at the temperature range of 23-30{degrees}C. This value of {alpha}{sub {omega}} is comparable to the data published by the IAEA. The {alpha}{sub {omega}} measured in the second core loading was found to be {minus}8 pcm/{degrees}C at the same temperature range. Another phenomenon considered in this study is core behavior during reactivity insertion transient. The results were compared to a core simulation using the Dynamic Simulator for Nuclear Power Plants. It was found that in the second core loading factors other than the moderator temperature influence the core reactivity more than expected. These effects proved to be extremely dependent on core configuration and may in certain core loadings render the reactor`s reactivity coefficient undesirable.

  2. Common Core State Standards for Students with Gifts and Talents

    Science.gov (United States)

    VanTassel-Baska, Joyce

    2015-01-01

    As many states have adopted the Common Core State Standards (CCSS), teachers can look to these standards as a framework for supporting students with gifts and talents. Differentiation of curriculum and instruction to address the CCSS will be necessary to meet the unique learning needs of learners with high ability and those with gifts and talents.…

  3. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  4. Application of nursing core competency standard education in the training of nursing undergraduates

    OpenAIRE

    Wu, Fang-qin; Wang, Yan-ling; Wu, Ying; Guo, Ming

    2014-01-01

    Purpose: To evaluate the effectiveness of nursing core competency standard education in undergraduate nursing training. Methods: Forty-two nursing undergraduates from the class of 2007 were recruited as the control group receiving conventional teaching methods, while 31 students from the class of 2008 were recruited as the experimental group receiving nursing core competency standard education. Teaching outcomes were evaluated using comprehensive theoretical knowledge examination and objec...

  5. Cognitive Language and Content Standards: Language Inventory of the Common Core State Standards in Mathematics and the Next Generation Science Standards

    Science.gov (United States)

    Winn, Kathleen M.; Mi Choi, Kyong; Hand, Brian

    2016-01-01

    STEM education is a current focus of many educators and policymakers and the Next Generation Science Standards (NGSS) with the Common Core State Standards in Mathematics (CCSSM) are foundational documents driving curricular and instructional decision making for teachers and students in K-8 classrooms across the United States. Thus, practitioners…

  6. FRG-1: new millenium - new compact core

    International Nuclear Information System (INIS)

    Schreiner, P.; Knop, W.

    2001-01-01

    The GKSS research center Geesthacht GmbH operates the MTR-type swimming pool research reactor FRG-1 (5 MW) for more than 40 years. The FRG-1 has been converted in February 1991 from HEU (93 %) to LEU (20 %) in one step and at that time the core size was reduced from 49 to 26 fuel elements. Consequently the thermal neutron flux in beam tube positions could be increased by more than a factor of two. It is the strong intention of GKSS to continue the operation of the FRG-1 research reactor for at least an additional 15 years with high availability and utilization. The reactor has been operated during the last years for approximately 250 full power days per year. To prepare the FRG-1 for an efficient future use, the core size has been reduced in a second step from 26 fuel elements to 12 fuel elements. (author)

  7. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  8. The common core mathematics standards transforming practice through team leadership

    CERN Document Server

    Hull, Ted H; Balka, Don S

    2012-01-01

    Transform math instruction with effective CCSS leadership The Common Core State Standards for mathematics describe the "habits of mind" that teachers should develop in their students without which the content standards cannot be successfully implemented. This professional development resource helps principals and math leaders grapple with the changes that must be addressed so that teachers can implement the practices required by the CCSS. Included are: A clear explanation of the CCSS for Mathematical Practice  Techniques to help leadership teams collaboratively implement and maintain the new standards A proficiency matrix with examples of instructional strategies for helping students reach competency in each standard.

  9. Unattended Monitoring of HEU Production in Gaseous Centrifuge Enrichment Plants using Automated Aerosol Collection and Laser-based Enrichment Assay

    International Nuclear Information System (INIS)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-01-01

    Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward low carbon energy production. Pivotal to the global nuclear power renaissance is the development and deployment of robust safeguards instrumentation that allows the limited resources of the IAEA to keep pace with the expansion of the nuclear fuel cycle. Undeclared production of highly enriched uranium (HEU) remains a primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs), due to their massive separative work unit (SWU) processing power and comparably short cascade equilibrium timescale. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely detection of HEU production within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. Our prior investigation demonstrated single-shot detection sensitivity approaching the femtogram range and relative isotope ratio uncertainty better than 10% using gadolinium as a surrogate for uranium. In this paper we present measurement results on standard samples containing traces of depleted, natural, and low enriched uranium, as well as measurements on aerodynamic size uranium particles mixed in background materials (e.g., dust, minerals, soils). Improvements and optimizations in the detection electronics, signal timing, calibration, and laser alignment have lead to significant improvements in detection sensitivity and enrichment accuracy, contributing to an overall reduction in the false alarm probability. The sample substrate media was also found to play a significant role in facilitating laser-induced vaporization and the production of energetic plasma conditions, resulting in ablation optimization and further improvements in the isotope abundance sensitivity.

  10. A detailed neutronics comparison of the university of Florida training reactor (UFTR) current HEU and proposed LEU cores

    International Nuclear Information System (INIS)

    Dionne, B.; Haghighat, A.; Yi, C.; Smith, R.; Ghita, G.; Manalo, K.; Sjoden, G.; Huh, J.; Baciak, J.; Mock, T.; Wenner, M.; Matos, J.; Stillman, J.

    2006-01-01

    For over 35 years, the UFTR highly-enriched core has been safely operated. As part of the Reduced Enrichment for Research and Test Reactors Program, the core is currently being converted to low-enriched uranium fuel. The analyses presented in this paper were performed to verify that, from a neutronic perspective, a proposed low-enriched core can be operated as safely and as effectively as the highly-enriched core. Detailed Monte Carlo criticality calculations are performed to determine: i) Excess reactivity for different core configurations, ii) Individual integral blade worth and shutdown margin, iii) Reactivity coefficients and kinetic parameters, and iv) Flux profiles and core six-factor formula parameters. (authors)

  11. Examining the Common Core State Standards in Agricultural Education

    Science.gov (United States)

    McKim, Aaron J.; Lambert, Misty D.; Sorensen, Tyson J.; Velez, Jonathan J.

    2015-01-01

    The Common Core State Standards (CCSS) represent a shift in the American education system. Included in the CCSS are opportunities for agriculture teachers to integrate math and English language arts content into their curriculum. Using the theory of planned behavior, we sought to identify Oregon agriculture teachers' attitudes, familiarity with,…

  12. Coupled fast-thermal core 'HERBE', as the benchmark experiment at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2003-10-01

    Validation of the well-known Monte Carlo code MCNP TM against measured criticality data for the coupled fast-thermal HERBE. System at the RB research reactor is shown in this paper. Experimental data are obtained for regular HERBE core and for the cases of controlled flooding of the neutron converter zone by heavy water. Earlier calculations of these criticality parameters, done by combination of transport and diffusion codes using 2D geometry model are also compared to new calculations carried out by the MCNP code in 3D geometry, applying new detailed 3D model of the HEU fuel slug, developed recently. Satisfactory agreements in comparison of the HERBE criticality calculation results with experimental data, in spite complex heterogeneous composition of the HERBE core, are obtained and confirmed that HERBE core could be used as a criticality benchmark for coupled fast-thermal core. (author)

  13. The development of core standards for editing in South Africa | Law ...

    African Journals Online (AJOL)

    The development of core standards for editing in South Africa. Melanie Ann Law. Abstract. South African editors2currently work within a highly unregulated industry. One factor contributing to this lack of regulation is the absence of clear standards that define the role of the editor and demarcate the tasks and skills required to ...

  14. Cloud-Based Collaborative Writing and the Common Core Standards

    Science.gov (United States)

    Yim, Soobin; Warschauer, Mark; Zheng, Binbin; Lawrence, Joshua F.

    2014-01-01

    The Common Core State Standards emphasize the integration of technology skills into English Language Arts (ELA) instruction, recognizing the demand for technology-based literacy skills to be college- and career- ready. This study aims to examine how collaborative cloud-based writing is used in in a Colorado school district, where one-to-one…

  15. A Deeper Glimpse into the National Core Arts Standards for General Music

    Science.gov (United States)

    Zaffini, Erin Dineen

    2018-01-01

    The National Core Arts Standards in general music provide some exciting possibilities for music growth and understanding among our students. For those of us who are still unsure of how to read the standards or implement them in our classrooms, the standards also present some challenges for music educators. This article provides a deeper look into…

  16. Improvement in operating characteristics resulting from the addition of FLIP fuel to a standard TRIGA core

    International Nuclear Information System (INIS)

    Randall, J.D.; Feltz, D.E.; Godsey, T.A.; Schumacher, R.F.

    1974-01-01

    To overcome problems associated with fuel burnup the Nuclear Science Center of Texas A and M University decided to convert from standard TRIGA fuel to FLIP-TRIGA fuel. FLIP fuel, which incorporates erbium as a burnable poison and is enriched to 70 percent in U-235, has a calculated lifetime of 9/MW-years. Due to limited funds a core was designed with a central region of 35 FLIP elements surrounded by 63 standard elements. Calculations indicated that the core excess and neutron fluxes were satisfactory, but no prediction was made of the improvements in core lifetime. The reactivity loss due to burnup for a standard core was measured to be 1.54 cents/MW-day. The addition of 35 FLIP fuel elements has reduced this value to approximately 0.5 cents/MW-day. The incorporation of FLIP fuel has, therefore, increased the lifetime of the core by a factor of three using fuel that is only 20 percent more expensive. The mixed core has other advantages as well. The power coefficient is less, the effect of xenon is less, and the fluxes in experimental facilities are higher. Thus, the mixed core has significant advantages over standard TRIGA fuel. (U.S.)

  17. Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R.; Villa, M.; Steinhauser, G.; Boeck, H. [Vienna University of Technology-Atominstitut (Austria)

    2011-07-01

    The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

  18. The State of State Standards--and the Common Core--in 2010

    Science.gov (United States)

    Carmichael, Sheila Byrd; Martino, Gabrielle; Porter-Magee, Kathleen; Wilson, W. Stephen

    2010-01-01

    This review of state English language arts (ELA) and mathematics standards is the latest in a series of Fordham evaluations dating back to 1997. It comes at a critical juncture, as states across the land consider adoption of the Common Core State Standards. These are the authors' major findings: (1) Based on their criteria, the Common Core…

  19. The CORE Community: Career and Technical Education Teachers' Perceptions of the Common Core State Standards

    Science.gov (United States)

    Stair, Kristin S.; Warner, Wendy J; Hock, Gaea; Conrad, Michelle; Levy, Natalie

    2016-01-01

    The Common Core State Standards (CCSS) have been adopted in 43 states within the U.S. However, Career and Technical Education (CTE) teachers are often unsure how their programs can successfully integrate CCSS. The purpose of this study was to understand how participants in a CCSS professional development project perceive the CCSS and how they are…

  20. Spectrophotometric Evaluation of Polyetheretherketone (PEEK as a Core Material and a Comparison with Gold Standard Core Materials

    Directory of Open Access Journals (Sweden)

    Bogna Stawarczyk

    2016-06-01

    Full Text Available This study investigated the colorimetric properties of different veneering materials on core materials. Standardized specimens (10 mm × 10 mm × 1.5 mm reflecting four core (polyetheretherketone (PEEK, zirconia (ZrO2, cobalt–chromium–molybdenum alloy (CoCrMo, and titanium oxide (TiO2; thickness: 1.5 mm and veneering materials (VITA Mark II, IPS e.max CAD, LAVA Ultimate and VITA Enamic, all in shade A3; thickness: 0.5, 1.0, 1.5 and 2 mm, respectively were fabricated. Specimens were superimposed to assemblies, and the color was determined with a spectrophotometer (CieLab-System or a chair-side color measurement device (VITA EasyShade, respectively. Data were analyzed using three-, two-, and one-way ANOVA, a Chi2-test, and a Wilson approach (p < 0.05. The measurements with EasyShade showed A2 for VITA Mark II, A3.5 for VITA Enamic, B2 for LAVA Ultimate, and B3 for IPS e.max CAD. LabE-values showed significant differences between the tested veneering materials (p < 0.001. CieLab-System and VITA EasyShade parameters of the different assemblies showed a significant impact of core (p < 0.001, veneering material (p < 0.001, and thickness of the veneering material (p < 0.001. PEEK as core material showed comparable outcomes as compared to ZrO2 and CoCrMo, with respect to CieLab-System parameters for each veneering material. The relative frequency of the measured VITA EasyShade parameters regarding PEEK cores also showed comparable results as compared to the gold standard CoCrMo, regardless of the veneering material used.

  1. Assessment of the technical specifications for a flip-standard TRIGA core

    International Nuclear Information System (INIS)

    Feltz, D.E.; Randall, J.D.

    1974-01-01

    The Technical Specifications for the Texas A and M University mixed, FLIP-Standard TRIGA core were the first submitted and approved under the draft version of Standard ANS-15.1. According to one AEC official these were the best Technical Specifications ever issued to a Research Reactor. The Technical Specifications are evaluated after operating under them for over seven months. (author)

  2. Assessment of the technical specifications for a flip-standard TRIGA core

    Energy Technology Data Exchange (ETDEWEB)

    Feltz, D E; Randall, J D [Texas A and M University (United States)

    1974-07-01

    The Technical Specifications for the Texas A and M University mixed, FLIP-Standard TRIGA core were the first submitted and approved under the draft version of Standard ANS-15.1. According to one AEC official these were the best Technical Specifications ever issued to a Research Reactor. The Technical Specifications are evaluated after operating under them for over seven months. (author)

  3. Standard Test Method for Water Absorption of Core Materials for Structural Sandwich Constructions

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This test method covers the determination of the relative amount of water absorption by various types of structural core materials when immersed or in a high relative humidity environment. This test method is intended to apply to only structural core materials; honeycomb, foam, and balsa wood. 1.2 The values stated in SI units are to be regarded as the standard. The inch-pound units given may be approximate. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  4. Student Reading Growth Illuminates the Common Core Text-Complexity Standard: Raising Both Bars

    Science.gov (United States)

    Williamson, Gary L.; Fitzgerald, Jill; Stenner, Jackson A.

    2014-01-01

    The Common Core State Standards (CCSS) establish a challenging text-complexity standard for all high school graduates to read at college and workplace text-complexity levels. We argue that implementation of the CCSS standard requires concurrent examination of historical student reading-growth trends. An example of a historical student average…

  5. Adapting to Change: Teacher Perceptions of Implementing the Common Core State Standards

    Science.gov (United States)

    Burks, Brooke A.; Beziat, Tara L. R.; Danley, Sheree; Davis, Kashara; Lowery, Holly; Lucas, Jessica

    2015-01-01

    The current research study looked at secondary teachers' (grades 6-12) perceptions of their preparedness to implement the Common Core State Standards as well as their feelings about the training they have or have not received related to implementing the standards. The problem: Many conflicting views exist among teachers, parents, and others…

  6. Incidence of tissue coring during transseptal catheterization when using electrocautery and a standard transseptal needle.

    Science.gov (United States)

    Greenstein, Eugene; Passman, Rod; Lin, Albert C; Knight, Bradley P

    2012-04-01

    The application of radiofrequency electrocautery to a standard, open-ended transseptal needle has been used to facilitate transseptal puncture (TSP). The purpose of this study was to determine the incidence of cardiac tissue coring when this technique is used. A model using excised swine hearts submerged in a saline-filled basin was developed to simulate TSP with electrocautery and a standard transseptal needle. Punctures were performed without the use of electrocautery and by delivering radiofrequency energy to the transseptal needle using a standard electrocautery pen at 3 target sites (fossa ovalis, non-fossa ovalis septum, and aorta). The tissue of the submerged heart was gently tented, and the needle was advanced on delivery of radiofrequency. The devices were retracted, and the needle was flushed in a collection basin. None of the TSPs without cautery caused tissue coring. For TSPs using electrocautery, the frequency of coring was at least 21% for any puncture permutation used in the study and averaged 37% at septal sites (Pelectrocautery and a standard open-ended Brockenbrough needle resulted in coring of the septal tissue in 35% of cases (33 of 96 punctures).

  7. Refurbishment, core conversion and safety analysis of Apsara reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K.; Sasidharan, K.; Sengupta, S. [Bhabha Atomic Research Centre, Mumbai (India)]. E-mail: nram@@apsara.barc.ernet.in

    1998-07-01

    Apsara, a 1 MWt pool type reactor using HEU fuel has been in operation at the Bhabha Atomic Research Centre, Trombay since 1956. In view of the long service period seen by the reactor it is now planned to carry out extensive refurbishment of the reactor with a view to extend its useful life. It is also proposed to modify the design of the reactor wherein the core will be surrounded by a heavy water reflector tank to obtain a good thermal neutron flux over a large radial distance from the core. Beam holes and the majority of the irradiation facilities will be located inside the reflector tank. The coolant flow direction through the core will be changed from the existing upward flow to downward flow. A delay tank, located inside the pool, is provided to facilitate decay of short lived radioactivity in the coolant outlet from the core in order to bring down radiation field in the operating areas. Analysis of various anticipated operational occurrences and accident conditions like loss of normal power, core coolant flow bypass, fuel channel blockage and degradation of primary coolant pressure boundary have been performed for the proposed design. Details of the proposed design modifications and the safety analyses are given in the paper. (author)

  8. Non-standard constraints within In-Core Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    Maldonado, G.I. [University of Cincinnati, P.O. Box 210072, Cincinnati, OH 45221-0072 (United States); Torres, C. [Comision Federal de Electricidad, Gestion de Combustible, Mexico, D.F. (Mexico); Marrote, G.N.; Ruiz U, V. [Global Nuclear Fuel, Americas, LLC, PO Box 780, M/C A16, Wilmington, NC28402 (United States)]. e-mail: Ivan.Maldonado@uc.edu

    2004-07-01

    Recent advancements in the area of nuclear fuel management optimization have been considerable and widespread. Therefore, it is not surprising that the design of today's nuclear fuel reloads can be a highly automated process that is often accompanied by sophisticated optimization software and graphical user interfaces to assist core designers. Most typically, among other objectives, optimization software seeks to maximize the energy efficiency of a fuel cycle while satisfying a variety of safety, operational, and regulatory constraints. Concurrently, the general industry trend continues to be one of pursuing higher generating capacity (i.e., power up-rates) alongside cycle length extensions. As these increasingly invaluable software tools and ambitious performance goals are pursued in unison, more aggressive core designs ultimately emerge that effectively minimize the margins to limits and, in some cases, may turn out less forgiving or accommodating to changes in underlying key assumptions. The purpose of this article is to highlight a few 'non-standard', though common constraints that can affect a BWR core design but which are often difficult, if not impossible, to implement into an automated setting. In a way, this article indirectly emphasizes the unique and irreplaceable role of the experienced designer in light of 'real life' situations. (Author)

  9. Non-standard constraints within In-Core Fuel Management

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Torres, C.; Marrote, G.N.; Ruiz U, V.

    2004-01-01

    Recent advancements in the area of nuclear fuel management optimization have been considerable and widespread. Therefore, it is not surprising that the design of today's nuclear fuel reloads can be a highly automated process that is often accompanied by sophisticated optimization software and graphical user interfaces to assist core designers. Most typically, among other objectives, optimization software seeks to maximize the energy efficiency of a fuel cycle while satisfying a variety of safety, operational, and regulatory constraints. Concurrently, the general industry trend continues to be one of pursuing higher generating capacity (i.e., power up-rates) alongside cycle length extensions. As these increasingly invaluable software tools and ambitious performance goals are pursued in unison, more aggressive core designs ultimately emerge that effectively minimize the margins to limits and, in some cases, may turn out less forgiving or accommodating to changes in underlying key assumptions. The purpose of this article is to highlight a few 'non-standard', though common constraints that can affect a BWR core design but which are often difficult, if not impossible, to implement into an automated setting. In a way, this article indirectly emphasizes the unique and irreplaceable role of the experienced designer in light of 'real life' situations. (Author)

  10. LEU fuel cycle analyses for the Belgian BR2 Research Reactor

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1988-01-01

    Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

  11. Mathematical Modeling, Sense Making, and the Common Core State Standards

    Science.gov (United States)

    Schoenfeld, Alan H.

    2013-01-01

    On October 14, 2013 the Mathematics Education Department at Teachers College hosted a full-day conference focused on the Common Core Standards Mathematical Modeling requirements to be implemented in September 2014 and in honor of Professor Henry Pollak's 25 years of service to the school. This article is adapted from my talk at this conference…

  12. FNR demonstration experiments Part II: Subcadmium neutron flux measurements

    International Nuclear Information System (INIS)

    Wehe, D.K.; King, J.S.

    1983-01-01

    The FNR HEU-LEU Demonstration Experiments include a comprehensive set of experiments to identify and quantify significant operational differences between two nuclear fuel enrichments. One aspect of these measurements, the subcadmium flux profiling, is the subject of this paper. The flux profiling effort has been accomplished through foil and wire activations, and by rhodium self-powered neutron detector (SPND) mappings. Within the experimental limitations discussed, the program to measure subcadmium flux profiles, lead to the following conclusions: (1) Replacement of a single fresh HEU element by a fresh LEU element at the center of an equilibrium HEU core produces a local flux depression. The ratio of HEU to LEU local flux is 1.19 ± .036, which is, well within experimental uncertainty, equal to the inverse of the U-235 masses for the two elements. (2) Whole core replacement of a large 38 element equilibrium HEU core by a fresh or nearly unburned LEU core reduces the core flux and raises the flux in both D 2 O and H 2 O reflectors. The reduction in the central core region is 40% to 10.0% for the small fresh 29 element LEU core, and 16% to 18% for a 31 element LEU core 482) with low average burnup 2 O reflector fluxes relative to core fluxes as measured by SPND with a fixed value of sensitivity, are in gross disagreement with the same flux ratios measured by Fe and Rh wire activations. Space dependent refinements of S are calculated to give some improvement in the discrepancy but the major part of the correction remains to be resolved

  13. The CORE Community: Career and Technical Education Teachers' Perceptions of the Common Core State Standards after a Professional Development Training

    Science.gov (United States)

    Stair, Kristin; Hock, Gaea; Warner, Wendy; Levy, Natalie; Conrad, Michelle

    2017-01-01

    Since the 1983 U.S Department of Education's report, "A Nation at Risk," various educational initiatives have been developed to support an increase in state standards and greater educational accountability (Liebtag, 2013). Despite opportunities to link Common Core State Standards (CCSS) and instructional curriculum, CTE teachers often…

  14. Representation and Analysis of Chemistry Core Ideas in Science Education Standards between China and the United States

    Science.gov (United States)

    Wan, Yanlan; Bi, Hualin

    2016-01-01

    Chemistry core ideas play an important role in students' chemistry learning. On the basis of the representations of chemistry core ideas about "substances" and "processes" in the Chinese Chemistry Curriculum Standards (CCCS) and the U.S. Next Generation Science Standards (NGSS), we conduct a critical comparison of chemistry…

  15. Fuel Management at the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

    2011-07-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

  16. IRT‑Sofia, HEU to LEU conversion: regulatory approval tasks solution overview

    International Nuclear Information System (INIS)

    Mitev, Mladen; Belousov, Sergey; Dimitrov, Dobromir

    2014-01-01

    The HEU to LEU conversion of the IRT–Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy of the Bulgarian Academy of Sciences was jointly studied with the Argonne National Laboratory as a part of the RERTR Programme. The main purpose of the collaboration consisted in accomplishment of safety analyses and preparation of documents used for regulatory approval tasks solution. The main steps and results which are fundamental for the preparation of IRT–Sofia Safety Analyses Report including Operating Limits and Conditions are presented in this paper. The documents prepared by INRNE in accordance with the European nuclear safety requirements and IAEA recommendations were submitted for approval to the Bulgarian Nuclear Regulatory Agency at the end of 2010. Key words: research reactor, safety analyses report, Nuclear Regulatory Agency

  17. Common Core State Standards for ELA/Literacy and Next Generation Science Standards: Convergences and Discrepancies Using Argument as an Example

    Science.gov (United States)

    Lee, Okhee

    2017-01-01

    As the Common Core State Standards (CCSS) for English language arts (ELA)/literacy and the Next Generation Science Standards (NGSS) highlight connections across subject areas, convergences and discrepancies come into view. As a prominent example, this article focuses on how the CCSS and the NGSS treat "argument," especially in Grades…

  18. HEU Holdup Measurements on 321-M A-Lathe

    International Nuclear Information System (INIS)

    Dewberry, R.A.

    2002-01-01

    The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) of the Savannah River Site to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the solid waste Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Three measurement systems were used to determine highly enriched uranium (HEU) holdup. This report covers holdup measurements on the A-Lathe that was used to machine uranium-aluminum-alloy (U-Al). Our results indicated that the lathe contained more than the limits stated in the Waste Acceptance Criteria (WAC) for the solid waste E-Area Vaults. Thus the lathe was decontaminated three times and assayed four times in order to bring the amounts of uranium to an acceptable content. This report will discuss the methodology, Non-Destructive Assay (NDA) measurements, and results of the U-235 holdup on the lathe

  19. Core conversion anaylses for the Portuguese Research Reactor

    International Nuclear Information System (INIS)

    Matos, J.E.; Stevens, J.G.; Feldman, E.E.; Stillman, J.A.; Dunn, F.E.; Kalimullah, K.; Marques, J.G.; Barradas, N.P.; Ramos, A.R.; Kling, A.

    2006-01-01

    Design and safety analyses are presented for conversion of the Portuguese Research Reactor (RPI) from the use of HEU fuel to the use of LEU fuel. The analyses were performed jointly by the RERTR Program at the Argonne National Laboratory (ANL) and the Instituto Tecnologico e Nuclear (ITN). The LEU fuel assembly design uses U 3 Si 2 -Al dispersion fuel with 4.8 g U/cm 3 and is very similar to the HEU fuel design. The results of neutronic studies, steady-state thermal-hydraulic analyses, accident analyses, and revisions to the Operating Limits and Conditions demonstrate that the RPI reactor can be operated safely with the new LEU fuel assemblies. Delivery of the LEU fuel is expected around the end of 2006, with conversion in early 2007. The HEU fuel is planned to be returned to the US in 2008.

  20. Why does the need of HEU for high flux research reactors remain?

    International Nuclear Information System (INIS)

    Glaeser, W.

    1991-01-01

    It has shown that high performance high flux reactors need an ongoing supply of highly enriched uranium. The new fuel materials in their highly enriched version offer prospective for advanced and better neutron sources vital for the future of neutron research. This is another very attractive result of the RERTR programme. One-sided restriction would only provide marginal or no values for research. If we adopt the sometimes expressed views that high enriched RERTR developed fuel should only be made available when unique benefits to mankind could be obtained, then certainly basic research at the forefront belongs to this category. HEU would only pose theoretical difficulties, if it would remain under proper safeguards and obviously this is the way to be pursued. (orig.)

  1. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    International Nuclear Information System (INIS)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

  2. A Research Agenda for the Common Core State Standards: What Information Do Policymakers Need?

    Science.gov (United States)

    Rentner, Diane Stark; Ferguson, Maria

    2014-01-01

    This report looks specifically at the information and data needs of policymakers related to the Common Core State Standards (CCSS) and the types of research that could provide this information. The ideas in this report were informed by a series of meetings and discussions about a possible research agenda for the Common Core, sponsored by the…

  3. A Comparison of Higher-Order Thinking between the Common Core State Standards and the 2009 New Jersey Content Standards in High School

    Science.gov (United States)

    Sforza, Dario; Tienken, Christopher H.; Kim, Eunyoung

    2016-01-01

    The creators and supporters of the Common Core State Standards claim that the Standards require greater emphasis on higher-order thinking than previous state standards in mathematics and English language arts. We used a qualitative case study design with content analysis methods to test the claim. We compared the levels of thinking required by the…

  4. Public Conceptions of Algorithms and Representations in the Common Core State Standards for Mathematics

    Science.gov (United States)

    Nanna, Robert J.

    2016-01-01

    Algorithms and representations have been an important aspect of the work of mathematics, especially for understanding concepts and communicating ideas about concepts and mathematical relationships. They have played a key role in various mathematics standards documents, including the Common Core State Standards for Mathematics. However, there have…

  5. 76 FR 30326 - Proposed Subsequent Arrangement

    Science.gov (United States)

    2011-05-25

    ... content of 28.276 kg (24.541 ekg) of U.S.-origin highly enriched uranium (HEU) (26.342 kg U-235) and 0.0048 g of plutonium contained in three HEU driver fuels that have been irradiated in the YAYOI nuclear... three HEU driver fuels from the core of YAYOI to be cut and de-cladded in the reactor room into...

  6. The Common Core State Standards and the Role of Instructional Materials: A Case Study on EdReports.org

    Science.gov (United States)

    Watt, Michael G.

    2016-01-01

    The purpose of this study was to review research studies investigating the role of instructional materials in relation to the Common Core State Standards and to evaluate whether a new organisation, EdReports.org, founded to evaluate the alignment of instructional materials to the Common Core State Standards, has achieved its objectives. Content…

  7. Safety Analysis Report for Packaging, Y-12 National Security Complex, Model ES-3100 Package with Bulk HEU Contents

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, James [Y-12 National Security Complex, Oak Ridge, TN (United States); Goins, Monty [Y-12 National Security Complex, Oak Ridge, TN (United States); Paul, Pran [Y-12 National Security Complex, Oak Ridge, TN (United States); Wilkinson, Alan [Y-12 National Security Complex, Oak Ridge, TN (United States); Wilson, David [Y-12 National Security Complex, Oak Ridge, TN (United States)

    2015-09-03

    This safety analysis report for packaging (SARP) presents the results of the safety analysis prepared in support of the Consolidated Nuclear Security, LLC (CNS) request for licensing of the Model ES-3100 package with bulk highly enriched uranium (HEU) contents and issuance of a Type B(U) Fissile Material Certificate of Compliance. This SARP, published in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guide 7.9 and using information provided in UCID-21218 and NRC Regulatory Guide 7.10, demonstrates that the Y-12 National Security Complex (Y-12) ES-3100 package with bulk HEU contents meets the established NRC regulations for packaging, preparation for shipment, and transportation of radioactive materials given in Title 10, Part 71, of the Code of Federal Regulations (CFR) [10 CFR 71] as well as U.S. Department of Transportation (DOT) regulations for packaging and shipment of hazardous materials given in Title 49 CFR. To protect the health and safety of the public, shipments of adioactive materials are made in packaging that is designed, fabricated, assembled, tested, procured, used, maintained, and repaired in accordance with the provisions cited above. Safety requirements addressed by the regulations that must be met when transporting radioactive materials are containment of radioactive materials, radiation shielding, and assurance of nuclear subcriticality.

  8. Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

    International Nuclear Information System (INIS)

    Hanan, N.A.; Liaw, J.R.; Matos, J.E.

    2005-01-01

    A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

  9. Quantitative Literacy and the Common Core State Standards in Mathematics

    Directory of Open Access Journals (Sweden)

    Bernard L. Madison

    2015-01-01

    Full Text Available How supportive of quantitative literacy (QL are the Common Core State Standards in Mathematics (CCSSM? The answer is tentative and conditional. There are some QL-supportive features including a strong probability and statistics strand in grade 6 through high school; a measurements and data strand in K-5; ratio and proportional reasoning standards in grades 6 and 7; and a comprehensive and coherent approach to algebraic reasoning and logical argument. However, the standards are weak in supporting reasoning and interpretation, and there are indications that the applications in CCSSM – mostly unspecified – will not include many QL contextual situations. Early indicators of assessment items follow a similar path. Except for statistics, most of the high school standards are aimed at development of algebra and precalculus topics, and there will likely be little room for more sophisticated applications of the QL-friendly mathematics of grades 6-8. The experience with CCSSM is limited at this point, leaving several crucial results uncertain, including assessments, emphases on statistics, and kinds of modeling and other applications.

  10. A Comparison of the American Common Core State Standards with the Finnish Educational System

    Science.gov (United States)

    Lynch, Kelly

    2014-01-01

    With the failure of the No Child Left Behind policies of the 1990's, educational reformers wished to establish a "new and improved" set of standards for the United States to follow. However, since their inception in 2006-2007, the new Common Core State Standards have become increasingly unpopular due to the fact that they remain largely…

  11. Alternative dispositioning methods for HEU spent nuclear fuel at the Savannah River Site

    International Nuclear Information System (INIS)

    Krupa, J.F.; McKibben, J.M.; Parks, P.B.; DuPont, M.E.

    1995-01-01

    The United States has a strong policy on prevention of the international spread of nuclear weapons. This policy was announced in Presidential Directive PDD-13 and summarized in a White House press release September 27, 1993. Two cornerstones of this policy are: seek to eliminate where possible the accumulation of stockpiles of highly- enriched uranium or plutonium; propose hor-ellipsis prohibiting the production of highly-enriched uranium (HEU) or plutonium for nuclear explosives purposes or outside international safeguards. The Department of Energy is currently struggling to devise techniques that safely and efficiently dispose of spent nuclear fuel (SNF) while satisfying national non-proliferation policies. SRS plans and proposals for disposing of their SNF are safe and cost effective, and fully satisfy non-proliferation objectives

  12. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  13. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    International Nuclear Information System (INIS)

    Montierth, Leland M.

    2016-01-01

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  14. Common Core Standards, Professional Texts, and Diverse Learners: A Qualitative Content Analysis

    Science.gov (United States)

    Yanoff, Elizabeth; LaDuke, Aja; Lindner, Mary

    2014-01-01

    This research study questioned the degree to which six professional texts guiding implementation of the Common Core Standards in reading address the needs of diverse learners. For the purposes of this research, diverse learners were specifically defined as above grade level readers, below grade level readers, and English learners. The researchers…

  15. Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

    Directory of Open Access Journals (Sweden)

    Xiaobo Liu

    2017-01-01

    Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

  16. A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

    International Nuclear Information System (INIS)

    Pond, R.B.; Hanan, N.A.; Matos, J.E.

    1997-01-01

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

  17. Integrating the English Language Arts Common Core State Standards into Physical Education

    Science.gov (United States)

    James, Alisa R.; Bullock, Kerri

    2015-01-01

    Physical education teachers are expected to implement the English language arts (ELA) Common Core State Standards (CCSS) in their instruction. This has proved to be challenging for many physical educators. The purpose of this article is to provide developmentally appropriate examples of how to incorporate the ELA CCSS into physical education,…

  18. FNR demonstration experiments Part I: Beam port leakage currents and spectra

    International Nuclear Information System (INIS)

    Wehe, D.K.; King, J.S.

    1983-01-01

    The goal of the NR-LEU experimental program has been to measure the changes in numerous reactor characteristics when the conventional HEU core is replaced by a complete LEU fueled core or by a single LEU element in the normal HEU core. We have observed comparisons in a) thermal flux intensity, spatial distribution and cadmium ratios, both in the core and in the light and heavy water reflectors, b) fast flux intensity and spectral shape at a special element within the core, c) the thermal leakage flux intensity at the exit positions of several beam ports and its spectral shape at one beam port, d) shim and control rod worths, e) temperature coefficient of reactivity, and f) xenon poison worth. The NR is a 2 MW light water pool reactor, reflected on three faces by light water and on one face by D 2 O, composed of MTR plate fuel elements. Figure shows a plan view of the core grid, D 2 O reflector tank, and beam ports. The conventional HEU fuel element contains eighteen MTR Al plates 30 in x 24 in x 0.06 in. The center 0.02 in of each plate is 93% U-235 enriched UAl x . A normal equilibrium HEU core loading is outlined. Each new HEU element contains ∼ 140 grams of U-235. The LEU low enrichment fuel retains the same plate and element geometry but the fuel is contained in a central 0.03 in thick UA l x matrix with 19.5% U-235 enrichment. Each new LEU element contains ov 167.3 grams U-235. In-core neutron fluxes were routinely mapped by a rhodium SPND and by many wire and foil activations. The same data, but in more restricted positions, were obtained through the light water reflector (south) and D 2 O reflector tank (north). Beam port leakage currents were measured during all power cycles, by transmission fission chambers at the exits of ports GI, and J, by a B3 detector at A-port, and by a prompt detector at the F-port exit. Thermal neutron spectra for both HEU and LEU cores were measured at I port using a single crystal silicon diffractometer. These measurements

  19. Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-10-01

    Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.

  20. Future Standardization of Space Telecommunications Radio System with Core Flight System

    Science.gov (United States)

    Briones, Janette C.; Hickey, Joseph P.; Roche, Rigoberto; Handler, Louis M.; Hall, Charles S.

    2016-01-01

    NASA Glenn Research Center (GRC) is integrating the NASA Space Telecommunications Radio System (STRS) Standard with the Core Flight System (cFS), an avionics software operating environment. The STRS standard provides a common, consistent framework to develop, qualify, operate and maintain complex, reconfigurable and reprogrammable radio systems. The cFS is a flexible, open architecture that features a plugand- play software executive called the Core Flight Executive (cFE), a reusable library of software components for flight and space missions and an integrated tool suite. Together, STRS and cFS create a development environment that allows for STRS compliant applications to reference the STRS application programmer interfaces (APIs) that use the cFS infrastructure. These APIs are used to standardize the communication protocols on NASAs space SDRs. The cFS-STRS Operating Environment (OE) is a portable cFS library, which adds the ability to run STRS applications on existing cFS platforms. The purpose of this paper is to discuss the cFS-STRS OE prototype, preliminary experimental results performed using the Advanced Space Radio Platform (ASRP), the GRC S- band Ground Station and the SCaN (Space Communication and Navigation) Testbed currently flying onboard the International Space Station (ISS). Additionally, this paper presents a demonstration of the Consultative Committee for Space Data Systems (CCSDS) Spacecraft Onboard Interface Services (SOIS) using electronic data sheets (EDS) inside cFE. This configuration allows for the data sheets to specify binary formats for data exchange between STRS applications. The integration of STRS with cFS leverages mission-proven platform functions and mitigates barriers to integration with future missions. This reduces flight software development time and the costs of software-defined radio (SDR) platforms. Furthermore, the combined benefits of STRS standardization with the flexibility of cFS provide an effective, reliable and

  1. A non-freaked out guide to teaching the common core using the 32 literacy anchor standards to develop college- and career-ready students

    CERN Document Server

    Stuart, Dave

    2014-01-01

    Implement the Common Core for ELA without all the stress A Non-Freaked Out Guide to Teaching the Common Core uses the often-neglected anchor standards to get to the heart of the Common Core State Standards (CCSS)-teaching students the skills they need to be college and career ready. Each anchor standard is broken down into its key points, and a discussion of each anchor standard''s central purpose helps outline the context for each required skill. This easy-to-read guide gives educators the kind of clear explanations, examples, and strategies they need to feel comfortable teaching the CCSS, an

  2. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    1983-09-01

    Separate abstracts have been prepared for each paper presented in the following areas of interest: (1) fuel development; (2) post-irradiation examinations; (3) reprocessing; (4) thermite reaction; (5) fuel fabrication; (6) element tests; (7) core tests; (8) criticals; (9) shipping; and (10) reactors and methods

  3. Investigating the Language Demands in the Common Core State Standards for English Language Learners: A Comparison Study of Standards

    Science.gov (United States)

    Wolf, Mikyung Kim; Wang, Yuan; Huang, Becky H.; Blood, Ian

    2014-01-01

    This study reports on a critical review of the language demands contained in the Common Core State Standards for English language arts (CCSS-ELA) with the aim of deriving important implications for the instruction of English language learners. The language demands of the CCSS-ELA were compared with those of existing English language arts (ELA) and…

  4. The Common Core State Standards' Quantitative Text Complexity Trajectory: Figuring out How Much Complexity Is Enough

    Science.gov (United States)

    Williamson, Gary L.; Fitzgerald, Jill; Stenner, A. Jackson

    2013-01-01

    The Common Core State Standards (CCSS) set a controversial aspirational, quantitative trajectory for text complexity exposure for readers throughout the grades, aiming for all high school graduates to be able to independently read complex college and workplace texts. However, the trajectory standard is presented without reference to how the…

  5. Integration of the Common Core State Standards into CTE: Challenges and Strategies of Career and Technical Teachers

    Science.gov (United States)

    Asunda, Paul A.; Finnell, Alicia M.; Berry, Nicholas R.

    2015-01-01

    In recent years, conversations about the importance of education standards in our school systems have intensified. Common Core State Standards (CCSS) are being implemented across most of the country. The standards require a major shift in instruction and the needed supports really are not there. This study investigated the common barriers,…

  6. Selection and benchmarking of computer codes for research reactor core conversions

    International Nuclear Information System (INIS)

    Yilmaz, E.; Jones, B.G.

    1983-01-01

    A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of Illinois. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general

  7. Teachers' Understanding of and Concerns about Mathematical Modeling in the Common Core Standards

    Science.gov (United States)

    Wolf, Nancy Butler

    2013-01-01

    Educational reform is most likely to be successful when teachers are knowledgeable about the intended reform, and when their concerns about the reform are understood and addressed. The Common Core State Standards (CCSS) is an effort to establish a set of nationwide expectations for students and teachers. This study examined teacher understanding…

  8. Experiments with HEU (93.14 wt.%) metal annuli with internal graphite cylinder

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaobo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wehmann, Udo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, only three experimental configurations are described here. They are internal graphite reflected metal uranium assemblies with three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches). These experiments can be found in Reference 1 and in their associated logbook

  9. Handwriting and Common Core State Standards: Teacher, Occupational Therapist, and Administrator Perceptions From New York State Public Schools.

    Science.gov (United States)

    Collette, Debra; Anson, Kylie; Halabi, Nora; Schlierman, April; Suriner, Allison

    Handwriting is the cornerstone of written performance and communication for school-age children. This mixed-methods study explored the impact of Common Core State Standards on handwriting instruction and its effects on perceptions regarding children's written responses in elementary school. Using surveys and interviews of elementary teachers, occupational therapists, and administrators in New York State public schools, we sought to understand current trends in handwriting instruction, changes in time spent on handwriting instruction in the classroom, supports offered to students who did not meet expectations for handwriting, and the impact of Common Core on children's written expression. Themes emerged revealing decreased handwriting instruction time and inconsistent use of handwriting instructional programs in the classroom after implementation of Common Core. Handwriting should be considered as a greater component in the foundational standards in Common Core. Occupational therapy services can support handwriting instruction implementation. Copyright © 2017 by the American Occupational Therapy Association, Inc.

  10. Developing standardized connection analysis techniques for slim hole core rod designs

    International Nuclear Information System (INIS)

    Fehr, G.; Bailey, E.I.

    1994-01-01

    Slim hole core rod design remains essentially in the proprietary domain. API standardization provides the ability to perform engineering analyses and dimensional inspections through the use of documents, ie: Specifications, Bulletins, and Recommended Practices. In order to provide similar engineering capability for non-API slim hole connections, this paper develops the initial phase of what may evolve into an engineering tool to provide at least an indication of relative serviceability between two connection styles for a given application. The starting point for this process will look at bending strength ratios and connection strength calculations. Since empirical data are yet needed to verify the approaches proposed in this paper, it is recognized that the alternatives presented here are only a first step to developing useful rules of thumb which may lead to later standardization

  11. The Relationship between Teacher Attitudes toward the Common Core State Standards and Informational Text

    Science.gov (United States)

    Estruch, Marcie Jane

    2018-01-01

    This study sought to determine the relationship between teachers' attitudes toward the Common Core State Standards and three predetermined factors. These factors were (1) teachers' attitudes toward the practicality of pedagogical shift three, balancing informational and literary texts, (2) teachers' attitudes toward school support with the…

  12. Full standard triple wireless transmission over 50m large core diameter graded index POF

    NARCIS (Netherlands)

    Shi, Y.; Morant, M.; Tangdiongga, E.; Llorente, R.; Koonen, A.M.J.

    2011-01-01

    We demonstrated for the first time a successful radio-over-1mm core diameter plastic optical fibre transmission of three simultaneous full standard wireless signals. Up to 50-m long transmission distance employing an eye-safe vertical cavity surface emitting laser has been achieved. The transmission

  13. Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Silva Macedo, L.V. da

    1989-01-01

    In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

  14. National Sexuality Education Standards: Core Content and Skills, K-12. A Special Publication of the Journal of School Health. Special Report

    Science.gov (United States)

    American School Health Association (NJ1), 2012

    2012-01-01

    The goal of this paper, "National Sexuality Education Standards: Core Content and Skills, K-12," is to provide clear, consistent and straightforward guidance on the "essential minimum, core content" for sexuality education that is developmentally and age-appropriate for students in grades K-12. The development of these standards is a result of an…

  15. Comment on the contribution of S.C. Mo, N.A. Hanan and J.E. Matos: 'Comparison of the FRM-II HEU design with an alternative LEU design'

    International Nuclear Information System (INIS)

    Boening, K.

    2004-01-01

    The results of the reference paper, which came to our attention for the first time during this RERTR Meeting, are more or less consistent with neutronic data we have obtained earlier within the FRM-II project (i.e. with own calculations and extrapolations). However, a realistic comparison of the HEU design of the FR.M-II (HEU = highly enriched uranium, 93 % U-235) with an alternative LEU design (LEU = low enriched uranium, 20 % U-235) is only possible on the basis of identical assumptions on the input parameters and has to consider more than neutronic data only. Serious scientists and experts should not confuse the politicians with academic studies touching some aspects of the full story only. The comparison has shown that the performance and reliability of the FRM-II design, which uses HEU fuel, is so advantageous that it can not - not even approximately - be met by an alternative design using LEU fuel. A change of the FRM-II design from HEU to LEU fuel with the results as shown above - i.e. less performance, higher costs, more nuclear waste and higher risk potential, and all of this with a delay of at least 5 years this could never be justified. If a future development of more advanced fuels should allow us to achieve our scientific goals at the conditions as identified above also with uranium of reduced enrichment - there would be no objection to a corresponding later conversion. Activities to realize a new neutron source in Germany go back to the late 70's with the project of a new middle flux beam reactor (MSR), which was abandoned shortly later in favour of an ambitious new spallation neutron source (SNQ). After this project also having been terminated around 1985 because of too high costs and technological risks, the hopes of the German community of neutron scientists focussed on the FRM-II. If non-technical pressure would damage this project this would equally provide irreversible damage to the large and still prospering field of neutron research in Germany

  16. Standardization of 32P activity determination method in soil-root cores for root distribution studies

    International Nuclear Information System (INIS)

    Sharma, R.B.; Ghildyal, B.P.

    1976-01-01

    The root distribution of wheat variety UP 301 was obtained by determining the 32 P activity in soil-root cores by two methods, viz., ignition and triacid digestion. Root distribution obtained by these two methods was compared with that by standard root core washing procedure. The percent error in root distribution as determined by triacid digestion method was within +- 2.1 to +- 9.0 as against +- 5.5 to +- 21.2 by ignition method. Thus triacid digestion method proved better over the ignition method. (author)

  17. Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

    1992-01-01

    The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

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

    International Nuclear Information System (INIS)

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

    1992-01-01

    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, each 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

  19. Michigan Senate Bill 826: Replace Common Core with pre-2011 Massachusetts Standards

    Directory of Open Access Journals (Sweden)

    Sandra Stotsky

    2016-04-01

    Full Text Available Interested in academic standards and assessments proven to raise student achievement? You won’t get that from the Common Core Standards and their associated consortium tests, PARCC and SBAC. Despite the boisterous hype of higher, deeper, richer, more rigorous, and so on, there exists no valid evidence to support their claims of higher quality, achievement, or college readiness. There is a set of state standards and assessments, however, proven through actual experience to have raised academic achievement for students at all levels and in all curricular pathways: those used in Massachusetts from 2000 to 2011. So, why not use them? Such a proposal was recently proposed, and passed, by the Michigan Senate Education Policy Committee. Here, we provide links to Sandra Stotsky’s testimony before that committee, along with other relevant links. - See more at: http://nonpartisaneducation.org/Review/Resources/MichiganBill.htm

  20. Strengthen Your Music Program by Incorporating Aspects of the ELA Common Core State Standards

    Science.gov (United States)

    Smith, Nancy Thompson

    2014-01-01

    Implementation of the English Language Arts (ELA) Common Core State Standards (CCSS) reduces the traditional separation between the study of different subjects. Increased focus on nonfiction reading and writing means more incorporation of other content, such as music, into language arts classes. CCSS's emphasis on speaking and writing across…

  1. Two-Dimensional Cutting (TDC Vitrectome: In Vitro Flow Assessment and Prospective Clinical Study Evaluating Core Vitrectomy Efficiency versus Standard Vitrectome

    Directory of Open Access Journals (Sweden)

    Mitrofanis Pavlidis

    2016-01-01

    Full Text Available Purpose. To evaluate comparative aspiration flow performance and also vitrectomy operating time efficiency using a double-cutting open port vitreous cutting system incorporated in a two-dimensional cutting (TDC, DORC International vitrectome design versus standard vitreous cutter. Methods. In vitro investigations compared aspiration flow rates in artificial vitreous humor at varying cutter speeds and vacuum levels using a TDC vitrectome and a standard vitrectome across different aspiration pump systems. A prospective single-centre clinical study evaluated duration of core vitrectomy in 80 patients with macular pucker undergoing 25-gauge or 27-gauge vitrectomy using either a TDC vitrectome at 16,000 cuts per minute (cpm or standard single-cut vitrectome, combined with a Valve Timing intelligence (VTi pump system (EVA, DORC International. Results. Aspiration flow rates remained constant independent of TDC vitrectome cut rate, while flow rates decreased linearly at higher cutter speeds using a classic single-blade vitrectome. Mean duration of core vitrectomy surgeries using a TDC vitreous cutter system was significantly (p<0.001 shorter than the mean duration of core vitrectomy procedures using a single-cut vitrectome of the same diameter (reduction range, 34%–50%. Conclusion. Vitrectomy surgery performed using a TDC vitrectome was faster than core vitrectomy utilizing a standard single-action vitrectome at similar cut speeds.

  2. RAFTing with Raptors: Connecting Science, English Language Arts, and the Common Core State Standards

    Science.gov (United States)

    Senn, Gary J.; McMurtrie, Deborah H.; Coleman, Bridget K.

    2013-01-01

    This article explores using the RAFT strategy (Role, Audience, Format, Topic) for writing in science classes. The framework of the RAFT strategy will be explained, and connections with Common Core State Standards (CCSS) for ELA/Literacy will be discussed. Finally, there will be a discussion of a professional learning experience for teachers in…

  3. Examining English Language Arts Common Core State Standards Instruction through Cultural Historical Activity Theory

    Science.gov (United States)

    Barrett-Tatum, Jennifer

    2015-01-01

    The English Language Arts Common Core State Standards and corresponding assessments brought about many changes for educators, their literacy instruction, and the literacy learning of their students. This study examined the day-to-day literacy instruction of two primary grade teachers during their first year of full CCSS implementation. Engestr?m's…

  4. Subcritical Neutron Multiplication Measurements of HEU Using Delayed Neutrons as the Driving Source

    International Nuclear Information System (INIS)

    Hollas, C.L.; Goulding, C.A.; Myers, W.L.

    1999-01-01

    A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutron detector using 3 He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations

  5. Leading Change for the Implementation of Common Core State Standards in Rural School Districts

    Science.gov (United States)

    Lopez, Paul; Wise, Donald

    2015-01-01

    Rural school districts across the nation, with their limited resources, face daunting challenges posed by the implementation of the Common Core State Standards. This article presents a recent study of 13 rural school districts in the Central Valley of California and how these districts are responding to those challenges. A total of 352 teachers…

  6. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  7. Using "The Joy Luck Club" to Teach Core Standards and 21st Century Literacies

    Science.gov (United States)

    Burns, Leslie David; Botzakis, Stergios G.

    2012-01-01

    In this article the authors illustrate an instructional unit based on a Common Core grades 9-10 illustrative text--Amy Tan's "Two Kinds" from "The Joy Luck Club." They demonstrate how teachers can meet the new standards "and" respond to students' 21st-century needs by using this modern classic along with other traditional and new media resources.…

  8. Examining Elementary Literacy Teachers' Perceptions of Their Preparedness to Implement the English Language Arts Common Core State Standards

    Science.gov (United States)

    Adams-Budde, Melissa; Miller, Samuel

    2015-01-01

    The purpose of our study was to examine elementary literacy teachers' perceptions of their preparedness to implement the ELA CCSS [English Language Arts Common Core State Standards]. We defined preparedness across three dimensions: teachers' perceived levels of knowledge of the standards and its components; efficacy to implement changes; and…

  9. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  10. Beyond the Core: Peer Observation Brings Common Core to Vocational and Electives Classes

    Science.gov (United States)

    Thurber Rasmussen, Harriette

    2014-01-01

    This article describes how a Washington State School District increased professional learning around the Common Core State Standards. The challenge was how to establish a way for career and technical education and electives teachers to learn and apply Common Core in their classes. Weaving Common Core literacy standards into vocational and…

  11. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  12. Overview of the Common Core State Standard initiative and educational reform movement from the vantage of speech-language pathologists.

    Science.gov (United States)

    Staskowski, Maureen

    2012-05-01

    Educational reform is sweeping the country. The adoption and the implementation of the Common Core State Standards in almost every state are meant to transform education. It is intended to update the way schools educate, the way students learn, and to ultimately prepare the nation's next generation for the global workplace. This article will describe the Common Core State Standard initiative and the underlying concerns about the quality of education in the United States as well as the opportunities this reform initiative affords speech-language pathologists. Thieme Medical Publishers 333 Seventh Avenue, New York, NY 10001, USA.

  13. HEU to LEU fuel conversion. Final report

    International Nuclear Information System (INIS)

    Mulder, R.U.

    1994-10-01

    The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG ampersand G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock ampersand Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B ampersand W) and the fuel designer (EG ampersand G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B ampersand W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology

  14. HEU to LEU fuel conversion. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.

    1994-10-01

    The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG&G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock & Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B&W) and the fuel designer (EG&G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B&W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  15. Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

    2012-10-01

    Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

  16. A neutronics study of LEU fuel options for the HFR-Petten

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1985-01-01

    The standard HEU fuel cycle characteristics are compared with those of several different LEU fuel cycles in the new vessel configuration. The primary design goals were to provide similar reactivity performance and neutron flux profiles with a minimal increase in 235 U loading. The fuel cycle advantages of Cd burnable absorbers over 10 B are presented. The LEU fuel cycle requirements were calculated also for an extended 32-day cycle and for a reload batch size reduction from six to five standard elements for the standard 26-day cycle. The effects of typical in-core experiments upon neutron flux profiles and fuel loading requirements are also presented. (author)

  17. Common Core State Standards for Mathematics: How Well Do the Textbook and Instructional Methods Align?

    Science.gov (United States)

    Rawding, Denise M.

    2016-01-01

    The Common Core Math Standards were written to address concerns that the math curriculum in the United States was not focused and coherent. Based on national and international assessments, the United States math scores have remained stagnant, while other countries have seen significant growth in their scores. This study, designed as an action…

  18. Integrating Apps with the Core Arts Standards in the 21st-Century Elementary Music Classroom

    Science.gov (United States)

    Heath-Reynolds, Julia; VanWeelden, Kimberly

    2015-01-01

    The implementation of the National Core Arts Standards has amplified the need for multiple approaches and opportunities for student responses and may compel music educators to use new tools. There are currently over one million available apps, and with the popularity of smart devices, student access to technology is increasing exponentially. Music…

  19. HEU Measurements of Holdup and Recovered Residue in the Deactivation and Decommissioning Activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    DEWBERRY, RAYMOND; SALAYMEH, SALEEM R.; CASELLA, VITO R.; MOORE, FRANK S.

    2005-03-11

    This paper contains a summary of the holdup and material control and accountability (MC&A) assays conducted for the determination of highly enriched uranium (HEU) in the deactivation and decommissioning (D&D) of Building 321-M at the Savannah River Site (SRS). The 321-M facility was the Reactor Fuel Fabrication Facility at SRS and was used to fabricate HEU fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the SRS production reactors. The facility operated for more than 35 years. During this time thousands of uranium-aluminum-alloy (U-Al) production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The bulk of this residue was located in the equipment that generated and handled small U-Al particles and in the exhaust systems for this equipment (e.g., Chip compactor, casting furnaces, log saw, lathes A & B, cyclone separator, Freon{trademark} cart, riser crusher, ...etc). The D&D project is likely to represent an important example for D&D activities across SRS and across the Department of Energy weapons complex. The Savannah River National Laboratory was tasked to conduct holdup assays to quantify the amount of HEU on all components removed from the facility prior to placing in solid waste containers. The U-235 holdup in any single component of process equipment must not exceed 50 g in order to meet the container limit. This limit was imposed to meet criticality requirements of the low level solid waste storage vaults. Thus the holdup measurements were used as guidance to determine if further decontamination of equipment was needed to ensure that the quantity of U-235 did not exceed the 50 g limit and to ensure that the waste met the Waste Acceptance Criteria (WAC) of the solid waste storage vaults. Since HEU is an accountable nuclear material, the holdup assays and assays of recovered

  20. A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).

    Energy Technology Data Exchange (ETDEWEB)

    Hanan, N. A.

    1998-01-14

    A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

  1. Long Range Active Detection of HEU Based on Thermal Neutron Multiplication

    Energy Technology Data Exchange (ETDEWEB)

    Forman L.; Dioszegi I.; Salwen, C.; and Vanier, P.E.

    2010-05-24

    We report on the results of measurements of proton irradiation on a series of targets at Brookhaven National Laboratory’s (BNL) Alternate Gradient Synchrotron Facility (AGS), in collaboration with LANL and SNL. We examined the prompt radiation environment in the tunnel for the DTRA-sponsored series (E 972), which investigated the penetration of air and subsequent target interaction of 4 GeV proton pulses. Measurements were made by means of an organic scintillator with a 500 MHz bandwidth system. We found that irradiation of a depleted uranium (DU) target resulted in a large gamma-ray signal in the 100-500 µsec time region after the proton flash when the DU was surrounded by polyethylene, but little signal was generated if it was surrounded by boron-loaded polyethylene. Subsequent Monte Carlo (MCNPX) calculations indicated that the source of the signal was consistent with thermal neutron capture in DU. The MCNPX calculations also indicated that if one were to perform the same experiment with a highly enriched uranium (HEU) target there would be a distinctive fast neutron yield in this 100-500 µsec time region from thermal neutron-induced fission. The fast neutrons can be recorded by the same direct current system and differentiated from gamma ray pulses in organic scintillator by pulse shape discrimination.

  2. Burn-up determinations and dimensional measurements of TRIGA-HEU fuel elements from the 14 MW steady-state core

    International Nuclear Information System (INIS)

    Toma, C.; Alexa, Al.; Craciunescu, T.; Pirvan, M.; Dobrin, R.

    2008-01-01

    In this paper there are presented the results of nondestructive examination in Post Irradiation Examination Laboratory for twenty five fuel rods selected from 14 MW steady state core. Gamma scanning and dimensional measurements were carried out in order to determine burn-up and diametric deflection of the fuel rods. Also, some comparisons with SSR Safety Report estimations for the maximum burn-up pin were made. (authors)

  3. Technical investigation of a pyrophoric event involving corrosion products from HEU ZPPR fuel plates

    International Nuclear Information System (INIS)

    Totemeier, T. C.

    2000-01-01

    A pyrophoric event recently occurred which involved corrosion products collected from highly-enriched uranium (HEU) fuel plates used in the Zero Power Physics Reactor (ZPPR). This paper summarizes the event and its background, and presents the results of an investigation into its source and mechanism. The investigation focused on characterization of corrosion product samples similar to those involved in the event using thermo-gravimetric analysis (TGA). Burning curve TGA tests were performed to measure the ignition temperature and hydride fractions of corrosion products in several different conditions to assess the effects of passivation treatment and long-term storage on chemical reactivity. The hydride fraction and ignition temperature of the corrosion products were found to be strongly dependent on the corrosion extent of the source metal. The results indicate that the energy source for the event was a considerable quantity of uranium hydride present in the corrosion products, but the specific ignition mechanism could not be identified

  4. Verification Results of Safety-grade Optical Modem for Core Protection Calculator (CPC) in Korea Standard Nuclear Power Plant (KSNP)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jangyeol; Son, Kwangseop; Lee, Youngjun; Cheon, Sewoo; Cha, Kyoungho; Lee, Jangsoo; Kwon, Keechoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    We confirmed that the coverage criteria for a safety-grade optical modem of a Core Protection Calculator is satisfactory using a traceability analysis matrix between high-level requirements and lower-level system test case data set. This paper describes the test environment, test components and items, a traceability analysis, and system tests as a result of system verification and validation based on Software Requirement Specifications (SRS) for a safety-grade optical modem of a Core Protection Calculator (CPC) in a Korea Standard Nuclear Power Plant (KSNP), and Software Design Specifications (SDS) for a safety-grade optical modem of a CPC in a KSNP. All tests were performed according to the test plan and test procedures. Functional testing, performance testing, event testing, and scenario based testing for a safety-grade optical modem of a Core Protection Calculator in a Korea Standard Nuclear Power Plant as a thirty-party verifier were successfully performed.

  5. An Analysis of the HEU-MET-FAST-035 Problem Using CENTRM and SCALE

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Jordan, W.C.

    1999-01-01

    An U/Fe benchmark, designated as HEU-MET-FAST-035, has been approved for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The SCALE code and cross sections performed poorly in calculating this critical experiment. Deficiencies in both the ENDF/B-V representation of the resonance region for Fe and in the Nordheim integral treatment when applied to Fe were identified. The combination of these deficiencies led to an almost 10% over-prediction of k(eff). Problems involving a large percentage of Fe and intermediate-energy spectrums present special cross-section processing difficulties for SCALE. In ENDF/B-V, resonance data for Fe only go to 400 keV, although resonances are present well above 1 MeV. Significant resonance data are stored as file 3 data instead of as resonance parameters. The Nordheim Integral Treatment used in NITAWL to process cross sections assume: resonances are widely spaced and all relevant information is contained in the resonance parameters (file 3 data is not processed). These limitations and assumptions result in poor solutions for this class of problems

  6. The MCART radiation physics core: the quest for radiation dosimetry standardization.

    Science.gov (United States)

    Kazi, Abdul M; MacVittie, Thomas J; Lasio, Giovanni; Lu, Wei; Prado, Karl L

    2014-01-01

    Dose-related radiobiological research results can only be compared meaningfully when radiation dosimetry is standardized. To this purpose, the National Institute of Allergy and Infectious Diseases (NIAID)-sponsored Medical Countermeasures Against Radiological Threats (MCART) consortium recently created a Radiation Physics Core (RPC) as an entity to assume responsibility of standardizing radiation dosimetry practices among its member laboratories. The animal research activities in these laboratories use a variety of ionizing photon beams from several irradiators such as 250-320 kVp x-ray generators, Cs irradiators, Co teletherapy machines, and medical linear accelerators (LINACs). In addition to this variety of sources, these centers use a range of irradiation techniques and make use of different dose calculation schemes to conduct their experiments. An extremely important objective in these research activities is to obtain a Dose Response Relationship (DRR) appropriate to their respective organ-specific models of acute and delayed radiation effects. A clear and unambiguous definition of the DRR is essential for the development of medical countermeasures. It is imperative that these DRRs are transparent between centers. The MCART RPC has initiated the establishment of standard dosimetry practices among member centers and is introducing a Remote Dosimetry Monitoring Service (RDMS) to ascertain ongoing quality assurance. This paper will describe the initial activities of the MCART RPC toward implementing these standardization goals. It is appropriate to report a summary of initial activities with the intent of reporting the full implementation at a later date.

  7. Qualification of high density aluminide fuels for the BR2 reactor

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, Andre; Gubel, Pol; Ponsard, Bernard; Pin, Thomas; Falgoux, Jean Louis

    2005-01-01

    The BR2 operation still relies on the use of 90..93% enriched HEU aluminide fuel. The availability of a limited batch of 73% enriched HEU from reprocessed BR2 uranium in Dounreay justified 10 years ago the qualification and use of this material. After some preliminary test irradiations, various batches of fuel elements were fabricated by the UKAEA-Dounreay and successfully irradiated. Due to their lower 235 U content (0.050 g 235 U/cm 2 ), these elements were always irradiated together with standard 90...93% HEU fuel elements. A mixed-core strategy was developed at this occasion for an optimal utilization, and was reported during the 4th RRFM conference (March 19-21, 2000, Colmar, France). The availability of a new batch of fresh 73% HEU material was the occasion, a few years ago, to initiate the development, fabrication and qualification of a new high density fuel element. An order was placed with CERCA to assess the optimal fabrication methods and tooling required to meet as far as possible the existing BR2 standard specifications and 235 U content (0.060 g 235 U/cm 2 ). This development phase has been already reported during the 7th RRFM conference (March 9-12, 2003, Aix-en-Provence, France). Afterwards, six lead test fuel elements were ordered for qualification by irradiation. The neutronic properties of the fuel elements were adjusted and optimized. After a short summary of the main results of the development program, this paper describes the nuclear characteristics of the high density fuel elements and comments on the nuclear follow-up of the lead test fuel elements during their irradiation for five cycles in the BR2 reactor and the return of experience for CERCA. (author)

  8. Modélisation et automatisation des procédés d’écriture et de production de supports de formation numérisés - Le modèle M.A.Ï.HEU.T.I.C. de la CCI de Paris

    Directory of Open Access Journals (Sweden)

    José Martin

    2005-01-01

    Full Text Available Cet article expose les conclusions d’un projet de recherche de la Direction de l’Enseignement de la Chambre de Commerce et d’Industrie de Paris (CCIP, mené de 2003 à 2005, dont l’aboutissement est un modèle didactique de production de contenus de cours numérisés, baptisé M.A.Ï.HEU.T.I.C.[[ M.A.Ï.HEU.T.I.C. : Modèle Appliqué d’Interprétation Heuristique des Technologies de l’Information et de la Communication.

  9. Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

    International Nuclear Information System (INIS)

    Dugan, E.T.; Diaz, N.J.; Kniedler, G.S.

    1983-01-01

    The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO 2 ) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of the

  10. Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dugan, E T; Diaz, N J [Department of Nuclear Engineering Sciences, University of Florida, Gainesville, FL (United States); Kniedler, G S [Reactor Analysis Group, TVA, Chattanooga, TN (United States)

    1983-09-01

    The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO{sub 2}) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of

  11. Incorporating the Common Core's Problem Solving Standard for Mathematical Practice into an Early Elementary Inclusive Classroom

    Science.gov (United States)

    Fletcher, Nicole

    2014-01-01

    Mathematics curriculum designers and policy decision makers are beginning to recognize the importance of problem solving, even at the earliest stages of mathematics learning. The Common Core includes sense making and perseverance in solving problems in its standards for mathematical practice for students at all grade levels. Incorporating problem…

  12. ZAKI: a windows-based ko standardization code for in-core INAA

    International Nuclear Information System (INIS)

    Ojo, J.O.; Filby, R.H.

    2002-01-01

    A new computer code ZAKI, for k o -based INAA standardization, written in Visual Basic for the WINDOWS environment is described. The parameter α measuring the deviation of the epithermal neutron spectrum shape from the ideal 1/E shape, and the thermal-to-epithermal flux ratio f, are monitored at each irradiation position for each irradiation using the ''triple bare monitor with k o '' technique. Stability of the irradiation position with respect to α and f is therefore assumed only for the duration of the irradiation. This now makes it possible to use k o standardization even for in-core reactor irradiation channels without an a priori knowledge of α and f values as required by existing commercial software. ZAKI is considerably versatile and contains features which allow for use of several detectors at different counting geometries, direct inputting of peak search output from GeniePc, and automatic nuclide identification of all gamma lines using an in-built library. Sample results for two certified reference materials are presented

  13. IC3 Internet and Computing Core Certification Global Standard 4 study guide

    CERN Document Server

    Rusen, Ciprian Adrian

    2015-01-01

    Hands-on IC3 prep, with expert instruction and loads of tools IC3: Internet and Computing Core Certification Global Standard 4 Study Guide is the ideal all-in-one resource for those preparing to take the exam for the internationally-recognized IT computing fundamentals credential. Designed to help candidates pinpoint weak areas while there's still time to brush up, this book provides one hundred percent coverage of the exam objectives for all three modules of the IC3-GS4 exam. Readers will find clear, concise information, hands-on examples, and self-paced exercises that demonstrate how to per

  14. Core Values | NREL

    Science.gov (United States)

    Core Values Core Values NREL's core values are rooted in a safe and supportive work environment guide our everyday actions and efforts: Safe and supportive work environment Respect for the rights physical and social environment Integrity Maintain the highest standard of ethics, honesty, and integrity

  15. Selection and benchmarking of computer codes for research reactor core conversions

    Energy Technology Data Exchange (ETDEWEB)

    Yilmaz, Emin [School of Aerospace, Mechanical and Nuclear Engineering, University of Oklahoma, Norman, OK (United States); Jones, Barclay G [Nuclear Engineering Program, University of IL at Urbana-Champaign, Urbana, IL (United States)

    1983-09-01

    A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC{sup 2}, COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k{sub eff} is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

  16. Selection and benchmarking of computer codes for research reactor core conversions

    International Nuclear Information System (INIS)

    Yilmaz, Emin; Jones, Barclay G.

    1983-01-01

    A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k eff is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

  17. The Common Core State Standards: An Opportunity to Enhance Formative Assessment in History/Social Studies Classrooms

    Science.gov (United States)

    Ateh, Comfort M.; Wyngowski, Aaron J.

    2015-01-01

    This article discusses the opportunity that the Common Core State Standards (CCSS) present for enhancing formative assessment (FA) in history and social studies classrooms. There is evidence that FA can enhance learning for students if implemented well. Unfortunately, teachers continue to be challenged in implementing FA in their classrooms. We…

  18. High School Teachers' Perspectives on the English Language Arts Common Core State Standards: An Exploratory Study

    Science.gov (United States)

    Ajayi, Lasisi

    2016-01-01

    This was an exploratory study that examined high school teachers' perspectives about their early experiences with the English language arts Common Core State Standards. The sources of data for the study included a survey and structured interviews. Twenty-three high school ELA teachers from one unified school district in Southern California…

  19. Common Core Standards and their Impact on Standardized Test Design

    NARCIS (Netherlands)

    Polleck, J.N.; Jeffery, J.V.

    2017-01-01

    With adoption of the Common Core (CCSS) in a majority of U.S. states came developmentof new high-stakes exams. Though researchers have investigated CCSS andrelated policies, less attention has been directed toward understanding how standardsare translated into testing. Due to the influence that

  20. Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-18

    This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met.

  1. Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1

    International Nuclear Information System (INIS)

    1995-01-01

    This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met

  2. Performance of PARR-1 with LEU Fuel

    International Nuclear Information System (INIS)

    Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

    2005-01-01

    Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

  3. Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1986-01-01

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235 U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m 3 . The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

  4. The Mailbox Computer System for the IAEA verification experiment on HEU downblending at the Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Aronson, A.L.; Gordon, D.M.

    2000-01-01

    IN APRIL 1996, THE UNITED STATES (US) ADDED THE PORTSMOUTH GASEOUS DIFFUSION PLANT TO THE LIST OF FACILITIES ELIGIBLE FOR THE APPLICATION OF INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA) SAFEGUARDS. AT THAT TIME, THE US PROPOSED THAT THE IAEA CARRY OUT A ''VERIFICATION EXPERIMENT'' AT THE PLANT WITH RESPECT TO DOOWNBLENDING OF ABOUT 13 METRIC TONS OF HIGHLY ENRICHED URANIUM (HEU) IN THE FORM OF URANIUM HEXAFLUROIDE (UF6). DURING THE PERIOD DECEMBER 1997 THROUGH JULY 1998, THE IAEA CARRIED OUT THE REQUESTED VERIFICATION EXPERIMENT. THE VERIFICATION APPROACH USED FOR THIS EXPERIMENT INCLUDED, AMONG OTHER MEASURES, THE ENTRY OF PROCESS-OPERATIONAL DATA BY THE FACILITY OPERATOR ON A NEAR-REAL-TIME BASIS INTO A ''MAILBOX'' COMPUTER LOCATED WITHIN A TAMPER-INDICATING ENCLOSURE SEALED BY THE IAEA

  5. Pollutant plume delineation from tree core sampling using standardized ranks

    DEFF Research Database (Denmark)

    Wahyudi, Agung; Bogaert, Patrick; Trapp, Stefan

    2012-01-01

    There are currently contradicting results in the literature about the way chloroethene (CE) concentrations from tree core sampling correlate with those from groundwater measurements. This paper addresses this issue by focusing on groundwater and tree core datasets in CE contaminated site, Czech...... Republic. Preliminary analyses revealed strongly and positively skewed distributions for the tree core dataset, with an intra-tree variability accounting for more than 80% of the total variability, while the spatial analyses based on variograms indicated no obvious spatial pattern for CE concentration...... groundwater and tree core measurements. Nonetheless, tree core sampling and analysis proved to be a quick and inexpensive semi-quantitative method and a useful tool....

  6. Fuel cycle cost comparisons with oxide and silicide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

    1983-09-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

  7. A seismic analysis of Korean standard PWR fuels under transition core conditions

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

    2005-01-01

    The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

  8. Constructing Assessment Model of Primary and Secondary Educational Quality with Talent Quality as the Core Standard

    Science.gov (United States)

    Chen, Benyou

    2014-01-01

    Quality is the core of education and it is important to standardization construction of primary and secondary education in urban (U) and rural (R) areas. The ultimate goal of the integration of urban and rural education is to pursuit quality urban and rural education. Based on analysing the related policy basis and the existing assessment models…

  9. The use of standardized patients in the plastic surgery residency curriculum: teaching core competencies with objective structured clinical examinations.

    Science.gov (United States)

    Davis, Drew; Lee, Gordon

    2011-07-01

    As of 2006, the Accreditation Council for Graduate Medical Education had defined six "core competencies" of residency education: interpersonal communication skills, medical knowledge, patient care, professionalism, practice-based learning and improvement, and systems-based practice. Objective structured clinical examinations using standardized patients are becoming effective educational tools, and the authors developed a novel use of the examinations in plastic surgery residency education that assesses all six competencies. Six plastic surgery residents, two each from postgraduate years 4, 5, and 6, participated in the plastic surgery-specific objective structured clinical examination that focused on melanoma. The examination included a 30-minute videotaped encounter with a standardized patient actor and a postencounter written exercise. The residents were scored on their performance in all six core competencies by the standardized patients and faculty experts on a three-point scale (1 = novice, 2 = moderately skilled, and 3 = proficient). Resident performance was averaged for each postgraduate year, stratified according to core competency, and scored from a total of 100 percent. Residents overall scored well in interpersonal communications skills (84 percent), patient care (83 percent), professionalism (86 percent), and practice-based learning (84 percent). Scores in medical knowledge showed a positive correlation with level of training (86 percent). All residents scored comparatively lower in systems-based practice (65 percent). The residents reported unanimously that the objective structured clinical examination was realistic and educational. The objective structured clinical examination provided comprehensive and meaningful feedback and identified areas of strengths and weakness for the residents and for the teaching program. The examination is an effective assessment tool for the core competencies and a valuable adjunct to residency training.

  10. Preservice Secondary Teachers' Conceptions from a Mathematical Modeling Activity and Connections to the Common Core State Standards

    Science.gov (United States)

    Stohlmann, Micah; Maiorca, Cathrine; Olson, Travis A.

    2015-01-01

    Mathematical modeling is an essential integrated piece of the Common Core State Standards. However, researchers have shown that mathematical modeling activities can be difficult for teachers to implement. Teachers are more likely to implement mathematical modeling activities if they have their own successful experiences with such activities. This…

  11. Stereotaxic core needle biopsy of breast microcalcifications obtained using a standard mammography table with an add-on unit

    International Nuclear Information System (INIS)

    Ward, S.E.; Taves, D.H.; McCurdy, L.I.

    2000-01-01

    To demonstrate the reliability of stereotaxic biopsy of indeterminate microcalcifications using a standard mammography table with an add-on unit. In 121 cases of indeterminate microcalcifications, core biopsy was performed using a standard mammography table with an add-on stereotaxic unit. Microcalcifications were identified on radiography of core specimens. Microcalcifications and a definitive histologic diagnosis were obtained in 112 core biopsies (92.6%), with no significant complications. In 23 lesions frank malignancy was diagnosed, and all of these diagnoses were confirmed on surgery. Pathologic examination suggested carcinoma in 4 lesions, and open biopsy confirmed malignancy in 3 of these cases. Four lesions showed atypical ductal hyperplasia. Benign disease was diagnosed in 81 lesions, of which 78 remained stable on mammographic follow-up (mean 16 months later) and 3 were subjected to surgical biopsy (of which 1 was malignant and 2 were benign). Nine cases were technically unsatisfactory because microcalcifications were not sampled. Stereotaxic core biopsy performed with an add-on unit is a safe and reliable technique for biopsy of indeterminate microcalcifications. For successful biopsy, microcalcifications must be harvested. Pathologic results should be correlated with mammographic findings. The accuracy rate compares favourably with results reported using prone biopsy tables. In an era of cost containment, this alternative to prone biopsy tables could result m significant savings in terms of capital investment and use of hospital rooms. In this study, surgical biopsy could have been avoided in 64.5% of cases. (author)

  12. Teachers' Perceptions on Preparedness and Supports to Implement the English Language Arts Common Core State Standards

    Science.gov (United States)

    Fernandez, Maria Clara

    2017-01-01

    The purpose of this study was to: (1) describe elementary teachers' perceptions on their preparedness to implement the English Language Arts Common Core State Standards (ELA-CCSS); (2) determine how perceptions influenced changes in instructional practices; and (3) to explore ELA-CCSS implementation challenges and/or barriers in supporting teacher…

  13. The Core Conversion of the TRIGA Reactor Vienna

    International Nuclear Information System (INIS)

    Villa, M.; Bergmann, R.; Musilek, A.; Sterba, J.H.; Böck, H.; Messick, C.

    2016-01-01

    The TRIGA Reactor Vienna has operated for many years with a mixed core using Al-clad and stainless-steel (SST) clad low enriched uranium (LEU) fuel and a few SST high enriched uranium (HEU) fuel elements. In view of the US spent fuel return program, the average age of these fuel elements and the Austrian position not to store any spent nuclear fuel on its territory, negotiation started in April 2011 with the US Department of Energy (DOE) and the International Atomic Energy Agency (IAEA). The sensitive subject was to return the old TRIGA fuel and to find a solution for a possible continuation of reactor operation for the next decades. As the TRIGA Vienna is the closest nuclear facility to the IAEA headquarters, high interest existed at the IAEA to have an operating research reactor nearby, as historically close cooperation exists between the IAEA and the Atominstitut. Negotiation started before summer 2011 between the involved Austrian ministries, the IAEA and the US DOE leading to the following solution: Austria will return 91 spent fuel elements to the Idaho National Laboratory (INL) while INL offers 77 very low burnt SST clad LEU elements for further reactor operation of the TRIGA reactor Vienna. The titles of these 77 new fuel elements will be transferred to Euratom in accordance with Article 86 of the Euratom-US Treaty. The fuel exchange with the old core returned to the INL, and the new core transferred to Vienna was carried out in one shipment in late 2012 through the ports of Koper/Slovenia and Trieste/Italy. This paper describes the administrative, logistic and technical preparations of the fuel exchange being unique world-wide and first of its kind between Austria and the USA performed successfully in early November 2012. (author)

  14. Portable optical frequency standard based on sealed gas-filled hollow-core fiber using a novel encapsulation technique

    DEFF Research Database (Denmark)

    Triches, Marco; Brusch, Anders; Hald, Jan

    2015-01-01

    A portable stand-alone optical frequency standard based on a gas-filled hollow-core photonic crystal fiber is developed to stabilize a fiber laser to the 13C2H2 P(16) (ν1 + ν3) transition at 1542 nm using saturated absorption. A novel encapsulation technique is developed to permanently seal...

  15. MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

    International Nuclear Information System (INIS)

    Sugino, K.; Iwai, T.

    2006-01-01

    MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

  16. Moving beyond Compliance: Promoting Research-Based Professional Discretion in the Implementation of the Common Core State Standards in English Language Arts

    Science.gov (United States)

    Woodard, Rebecca; Kline, Sonia

    2015-01-01

    State- and local-level mandates are currently being implemented to ensure strict compliance to the new national Common Core State Standards for English Language Arts (CCSS for ELA) and related assessments. These standards provide many potential opportunities to improve literacy education nationally and locally. However, the CCSS for ELA will…

  17. ZAKI a windows-based k sub o standardization code for in-core INAA

    CERN Document Server

    Ojo, J O

    2002-01-01

    A new computer code ZAKI, for k sub o -based INAA standardization, written in Visual Basic for the WINDOWS environment is described. The parameter alpha measuring the deviation of the epithermal neutron spectrum shape from the ideal 1/E shape, and the thermal-to-epithermal flux ratio f, are monitored at each irradiation position for each irradiation using the ''triple bare monitor with k sub o '' technique. Stability of the irradiation position with respect to alpha and f is therefore assumed only for the duration of the irradiation. This now makes it possible to use k sub o standardization even for in-core reactor irradiation channels without an a priori knowledge of alpha and f values as required by existing commercial software. ZAKI is considerably versatile and contains features which allow for use of several detectors at different counting geometries, direct inputting of peak search output from GeniePc, and automatic nuclide identification of all gamma lines using an in-built library. Sample results for ...

  18. Evaluation of remote monitoring at the Oak Ridge HEU storage vault -- First thoughts and final application

    International Nuclear Information System (INIS)

    Sheely, K.B.; Whitaker, J.M.

    1996-01-01

    Remote monitoring provides a more timely and comprehensive way to meet national and international requirements for monitoring nuclear material inventories. Unattended monitoring technologies could be used to meet national needs for nuclear material safety, protection, control and accountability. Unattended systems possessing a remote data transmission capability could be used to meet international requirements for nuclear material safeguards and transparency. Even though more enhancements are required to improve system reliability, remote monitoring''s future potential seems great. The key questions are: (1) how will remote monitoring systems be used (configuration and operation); (2) how effective will the system be (vs. current activities); and (3) how much will it cost. This paper provides preliminary answers to these questions based on the experience gained from a joint IAEA-United States Support Program (USSP) task to evaluate remote monitoring at the Oak Ridge HEU Storage vault. This paper also draws on experience gained from US involvement in other remote monitoring projects

  19. Preservice Secondary Teachers Perceptions of College-Level Mathematics Content Connections with the Common Core State Standards for Mathematics

    Science.gov (United States)

    Olson, Travis A.

    2016-01-01

    Preservice Secondary Mathematics Teachers (PSMTs) were surveyed to identify if they could connect early-secondary mathematics content (Grades 7-9) in the Common Core State Standards for Mathematics (CCSSM) with mathematics content studied in content courses for certification in secondary teacher preparation programs. Respondents were asked to…

  20. Current status of operation and utilization of the Dalat Research Reactor

    International Nuclear Information System (INIS)

    Dien, Nguyen Nhi

    2006-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool-type reactor using the HEU (36% enrichment) WWR-M2 fuel assemblies. It was renovated and upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in the 1st November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analysis, scientific research and training. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for reactor physics and thermal hydraulics experiments. From the first start-up to the end of December 2004, it totaled about 27,253 hrs of operation and the total energy released was about 543 MWd. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 fuel assemblies (FA). The 11 new FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 100 FAs. The second fuel reloading was executed in March 2002. The 4 new FAs were added in the core periphery, at previous beryllium element locations. The working configuration of 104 FAs ensured efficient exploitation of the DNRR at nominal power for about 3000 hrs since March 2002. In order to provide excess reactivity for the reactor operation without the need to discharge high burned FAs, in June 2004, the fuel shuffling of the reactor core was done. 16 FAs with low burn-up from the core periphery were moved toward the core center and 16 FAs with high-burn-up from the core center were moved toward the core periphery. This operation provided additional reactivity of about 0.85 β eff that the current reactor configuration using re-shuffled HEU fuel is expected to allow normal operation until June 2006. In 1999, the request of returning to Russia HEU fuels from foreign

  1. Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

    International Nuclear Information System (INIS)

    Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

    2000-09-01

    A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

  2. Conversion of the University of Missouri-Rolla Reactor from high-enriched uranium to low-enriched uranium fuel

    International Nuclear Information System (INIS)

    Bolon, A.E.; Straka, M.; Freeman, D.W.

    1997-01-01

    The objectives of this project were to convert the UMR Reactor fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel and to ship the HEU fuel back to the Department of Energy Savannah River Site. The actual core conversion was completed in the summer of 1992. The HEU fuel was offloaded to an onsite storage pit where it remained until July, 1996. In July, 1996, the HEU fuel was shipped to the DOE Savannah River Site. The objectives of the project have been achieved. DOE provided the following funding for the project. Several papers were published regarding the conversion project and are listed in the Attachment. In retrospect, the conversion project required much more time and effort than originally thought. Several difficulties were encountered including the unavailability of a shipping cask for several years. The authors are grateful for the generous funding provided by DOE for this project but wish to point out that much of their efforts on the conversion project went unfunded

  3. Neutronic analysis for conversion of the Ghana Research Reactor-1 facility using Monte Carlo methods and UO{sub 2} LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Akaho, E.H.K.; Maakuu, B.T.; Gbadago, J.K. [Ghana Research Reactor-1 Centre, Dept. of Nuclear Engineering and Materials Science, National Nuclear Research Institute, Ghana Atomic Energy Commission, Legon, Accra (Ghana); Andam, A. [Kwame Nkrumah Univ. of Science and Technology, Dept. of Physics (Ghana); Liaw, J.J.R.; Matos, J.E. [Argonne National Lab., RERTR Programme, Div. of Nuclear Engineering (United States)

    2007-07-01

    Monte Carlo particle transport methods and software (MCNP) have been applied to the modelling, simulation and neutronic analysis for the conversion of the HEU-fuelled (high enrichment uranium) core of the Ghana Research Reactor-1 (GHARR-1) facility. The results show that the MCNP model of the GHARR-1 facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR) is good as the simulated neutronic and other reactor physics parameters agree with very well with experimental and zero power results. Three UO{sub 2} LEU (low enrichment uranium) fuels with different enrichments (12.6% and 19.75%), core configurations, core loadings were utilized in the conversion studies. The nuclear criticality and kinetic parameters obtained from the Monte Carlo simulation and neutronic analysis using three UO{sub 2} LEU fuels are in close agreement with results obtained for the reference 90.2% U-Al HEU core. The neutron flux variation in the core, fission chamber and irradiation channels for the LEU UO{sub 2} fuels show the same trend as the HEU core as presented in the paper. The Monte Carlo model confirms a reduction (8% max) in the peak neutron fluxes simulated in the irradiation channels which are utilized for experimental and commercial activities. However, the reductions or 'losses' in the flux levels neither affects the criticality safety, reactor operations and safety nor utilization of the reactor. Employing careful core loading optimization techniques and fuel loadings and enrichment, it is possible to eliminate the apparent reductions or 'losses' in the neutron fluxes as suggested in this paper. Concerning neutronics, it can be concluded that all the 3 LEU fuels qualify as LEU candidates for core conversion of the GHARR-1 facility.

  4. Common Core State Standards for Literacy in History/Social Studies, Science, and Technical Subjects for English Language Learners

    Science.gov (United States)

    de Oliveira, Luciana C., Ed.

    2016-01-01

    This volume in the Common Core State Standards (CCSS) for English Language Learners series was designed to deepen teacher's knowledge and provides instructional approaches and practices for supporting grades 6-12 ELLs as they meet the ambitious expectations of the CCSS for Literacy in History/Social Studies, Science, and Technical Subjects. This…

  5. Writing to the Common Core: Teachers' Responses to Changes in Standards and Assessments for Writing in Elementary Schools

    Science.gov (United States)

    Wilcox, Kristen Campbell; Jeffery, Jill V.; Gardner-Bixler, Andrea

    2016-01-01

    This multiple case study investigated how the Common Core State Standards (CCSS) for writing and teacher evaluation system based in part on CCSS assessments might be influencing writing instruction in elementary schools. The sample included nine schools: Six achieved above-predicted performance on English Language Arts (ELA) as well as prior ELA…

  6. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels guidebook

    International Nuclear Information System (INIS)

    1980-08-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this document has been prepared to assist reactor operators in determining whether conversion to the use of low enriched uranium (LEU) fuel designs is technically feasible for their specific reactor, and to assist in making a smooth transition to the use of LEU fuel designs where appropriate

  7. Calibration Tools for Measurement of Highly Enriched Uranium in Oxide and Mixed Uranium-Plutonium Oxide with a Passive-Active Neutron Drum Shuffler

    International Nuclear Information System (INIS)

    Mount, M; O'Connell, W; Cochran, C; Rinard, P

    2003-01-01

    Lawrence Livermore National Laboratory (LLNL) has completed an extensive effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. Earlier papers described the PAN shuffler calibration over a range of item properties by standards measurements and an extensive series of detailed simulation calculations. With a single normalization factor, the simulations agree with the HEU oxide standards measurements to within ±1.2% at one standard deviation. Measurement errors on mixed U-Pu oxide samples are in the ±2% to ±10% range, or ±20 g for the smaller items. The purpose of this paper is to facilitate transfer of the LLNL procedure and calibration algorithms to external users who possess an identical, or equivalent, PAN shuffler. Steps include (1) measurement of HEU standards or working reference materials (WRMs); (2) MCNP simulation calculations for the standards or WRMs and a range of possible masses in the same containers; (3) a normalization of the calibration algorithms using the standard or WRM measurements to account for differences in the 252 Cf source strength, the delayed-neutron nuclear data, effects of the irradiation protocol, and detector efficiency; and (4) a verification of the simulation series trends against like LLNL results. Tools include EXCEL/Visual Basic programs which pre- and post-process the simulations, control the normalization, and embody the calibration algorithms

  8. The Role of District Office Leaders in the Adoption and Implementation of the Common Core State Standards in Elementary Schools

    Science.gov (United States)

    Durand, Francesca T.; Lawson, Hal A.; Wilcox, Kristen Campbell; Schiller, Kathryn S.

    2016-01-01

    Purpose: This multiple case study investigated district leaders' orientations and strategies as their elementary schools proceeded with state-mandated implementation of the Common Core State Standards (CCSS). We identified differences between schools achieving above-predicted outcomes on state CCSS assessments ("odds-beaters") and…

  9. DOE Radiological Control Manual Core Training Program

    International Nuclear Information System (INIS)

    Scott, H.L.; Maisler, J.

    1993-01-01

    Over the past year, the Department of Energy (DOE) Office of Health (EH-40) has taken a leading role in the development of new standardized radiological control training programs for use throughout the DOE complex. The Department promulgated its Radiological Control (RadCon) Manual in June 1992. To ensure consistent application of the criteria presented in the RadCon Manual, standardized radiological control core training courses and training materials have been developed for implementation at all DOE facilities. In producing local training programs, standardized core courses are to be supplemented with site-specific lesson plans, viewgraphs, student handbooks, qualification standards, question banks, and wallet-sized training certificates. Training programs for General Employee Radiological Training, Radiological Worker I and II Training, and Radiological Control Technician Training have been disseminated. Also, training committees under the direction of the Office of Health (EH-40) have been established for the development of additional core training courses, development of examination banks, and the update of the existing core training courses. This paper discusses the current activities and future direction of the DOE radiological control core training program

  10. Autopsy consent, brain collection, and standardized neuropathologic assessment of ADNI participants: the essential role of the neuropathology core.

    Science.gov (United States)

    Cairns, Nigel J; Taylor-Reinwald, Lisa; Morris, John C

    2010-05-01

    Our objectives are to facilitate autopsy consent, brain collection, and perform standardized neuropathologic assessments of all Alzheimer's Disease Neuroimaging Initiative (ADNI) participants who come to autopsy at the 58 ADNI sites in the USA and Canada. Building on the expertise and resources of the existing Alzheimer's Disease Research Center (ADRC) at Washington University School of Medicine, St. Louis, MO, a Neuropathology Core (NPC) to serve ADNI was established with one new highly motivated research coordinator. The ADNI-NPC coordinator provides training materials and protocols to assist clinicians at ADNI sites in obtaining voluntary consent for brain autopsy in ADNI participants. Secondly, the ADNI-NPC maintains a central laboratory to provide uniform neuropathologic assessments using the operational criteria for the classification of AD and other pathologies defined by the National Alzheimer Coordinating Center (NACC). Thirdly, the ADNI-NPC maintains a state-of-the-art brain bank of ADNI-derived brain tissue to promote biomarker and multi-disciplinary clinicopathologic studies. During the initial year of funding of the ADNI Neuropathology Core, there was notable improvement in the autopsy rate to 44.4%. In the most recent year of funding (September 1(st), 2008 to August 31(st) 2009), our autopsy rate improved to 71.5%. Although the overall numbers to date are small, these data demonstrate that the Neuropathology Core has established the administrative organization with the participating sites to harvest brains from ADNI participants who come to autopsy. Within two years of operation, the Neuropathology Core has: (1) implemented a protocol to solicit permission for brain autopsy in ADNI participants at all 58 sites who die and (2) to send appropriate brain tissue from the decedents to the Neuropathology Core for a standardized, uniform, and state-of-the-art neuropathologic assessment. The benefit to ADNI of the implementation of the NPC is very clear

  11. An alternative LEU design for the FRM-II

    International Nuclear Information System (INIS)

    Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

    1997-02-01

    The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm[sup 3] and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 10[sup 14] n/cm[sup 2]/s in the reflector). LEU silicide fuel with 4.5 g/cm[sup 3] has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Computer models for the HEU and LEU designs have been exchanged between TUM and ANL and discrepancies have been resolved. The following issues are addressed: qualification of HEU and LEU silicide fuels, stability of the fuel plates, gamma heating in the heavy water reflector, a hypothetical accident involving the configuration of the reflector, a loss of primary coolant flow transient due to an interrupted power supply, the radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. Calculations were also done to address the possibility that new high density LEU fuels could be developed that would allow conversion of the TUM HEU design to LEU fuel. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

  12. Irradiation of MEU and LEU test fuel elements in DR 3

    International Nuclear Information System (INIS)

    Haack, K.

    1984-01-01

    Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

  13. Common Core State Standards for English Language Arts & Literacy in History/Social Studies, Science, and Technical Subjects. Appendix C: Samples of Student Writing

    Science.gov (United States)

    Common Core State Standards Initiative, 2010

    2010-01-01

    This document presents writing samples that have been annotated to illustrate the criteria required to meet the Common Core State Standards for particular types of writing--argument, informative/explanatory text, and narrative--in a given grade. Each of the samples exhibits at least the level of quality required to meet the Writing standards for…

  14. Teaching to Exceed the English Language Arts Common Core State Standards: A Literacy Practices Approach for 6-12 Classrooms

    Science.gov (United States)

    Beach, Richard; Thein, Amanda Haertling; Webb, Allen

    2012-01-01

    As the new English Language Arts Common Core State Standards take hold across the United States, the need grows for pre-service and in-service teachers to be ready to develop curriculum and instruction that addresses their requirements. This timely, thoughtful, and comprehensive text directly meets this need. It delineates a literacy practices and…

  15. The Impact of Climatological Conditions on Low Enriched Uranium Loading Station Operations for the HEU Blend Down Project

    International Nuclear Information System (INIS)

    Chang, R.C.

    2002-01-01

    A computer model was developed using COREsim to perform a time motion study for the Low Enriched Uranium (LEU) Loading Station operations. The project is to blend Highly Enriched Uranium (HEU) with Natural Uranium (NU) to produce LEU to be shipped to Tennessee Valley Authority (TVA) for further processing. To cope with a project cost reduction, the LEU Loading Station concept has changed from an enclosed building with air-conditioning to a partially enclosed building without air conditioning. The LEU Loading Station is within a radiological contaminated area; two pairs of coveralls and negative pressure respirator are required. As a result, inclement weather conditions, especially heat stress, will affect and impact the LEU loading operations. The purposes of the study are to determine the climatological impacts on LEU Loading operations, resources required for committed throughputs, and to find out the optimum process pathways for multi crews working simultaneously in the space-lim ited LEU Loading Station

  16. A programme for Euratom safeguards inspectors, used in the assay of high enriched (H.E.U.) and low enriched (L.E.U.) uranium fuel materials by active neutron interrogation

    International Nuclear Information System (INIS)

    Vocino, V.; Farese, N.; Maucq, T.; Nebuloni, M.

    1991-01-01

    The programme AECC (Active Euratom Coincidence Counters) has been developed at the Joint Research Center, Ispra by the Euratom Safeguards Directorate, Luxembourg and the Safety Technology Institute, Ispra for the acquisition, evaluation, management and storage of measurement data originating from active neutron interrogation of HEU and LEU fuel materials. The software accommodates the implementation of the NDA (Non Destructive Assay) procedures for the Active Well Coincidence Counters and Active Neutron Coincidence Counters deployed by the Euratom Safeguards Directorate, Luxembourg

  17. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

  18. Production, inventories and HEU in the world uranium market: Production's vital role

    International Nuclear Information System (INIS)

    Underhill, D.H.

    1997-01-01

    This paper analyses recent uranium supply and demand relationship and projects supply through 2010. The extremely depressed record low market prices have led to the ongoing annual inventory drawdown of over 25,000 t U resulting from the current 45% world production shortfall. The policy of the European Union and anti-dumping related activities in the USA are restricting imports of uranium from CIS producers to a majority of the world's nuclear utilities. These factors are reducing low priced uranium supply and forcing buyers to again obtain more of their requirements from producers. It discusses how the sale of Low Enriched Uranium (LEU) produced from of 550 t High Enriched Uranium (HEU) from Russia and Ukraine could potentially supply about 15% of world requirements through 2010. However, legislation currently being developed by the US Congress may ration the sale of this material, extending the LEU supply well into the next century. Nuclear generation capacity and its uranium requirements are projected to grow at about 1.5% through 2010. Demand for new uranium purchases is however, increasing at the much higher rate of 25-30% over the next 10-15 years. This increasing demand in the face of decreasing supply is resulting in a market recovery in which the spot price for non-CIS produced uranium has risen over 25% since October 1994. Prices will continue to increase as the market equilibrium shifts from a balance with alternative excess low priced supply to an equilibrium between production and demand. 19 refs, 14 figs, 2 tabs

  19. Establishing Core Outcome Domains in Hemodialysis: Report of the Standardized Outcomes in Nephrology-Hemodialysis (SONG-HD) Consensus Workshop.

    Science.gov (United States)

    Tong, Allison; Manns, Braden; Hemmelgarn, Brenda; Wheeler, David C; Evangelidis, Nicole; Tugwell, Peter; Crowe, Sally; Van Biesen, Wim; Winkelmayer, Wolfgang C; O'Donoghue, Donal; Tam-Tham, Helen; Shen, Jenny I; Pinter, Jule; Larkins, Nicholas; Youssouf, Sajeda; Mandayam, Sreedhar; Ju, Angela; Craig, Jonathan C

    2017-01-01

    Evidence-informed decision making in clinical care and policy in nephrology is undermined by trials that selectively report a large number of heterogeneous outcomes, many of which are not patient centered. The Standardized Outcomes in Nephrology-Hemodialysis (SONG-HD) Initiative convened an international consensus workshop on November 7, 2015, to discuss the identification and implementation of a potential core outcome set for all trials in hemodialysis. The purpose of this article is to report qualitative analyses of the workshop discussions, describing the key aspects to consider when establishing core outcomes in trials involving patients on hemodialysis therapy. Key stakeholders including 8 patients/caregivers and 47 health professionals (nephrologists, policymakers, industry, and researchers) attended the workshop. Attendees suggested that identifying core outcomes required equitable stakeholder engagement to ensure relevance across patient populations, flexibility to consider evolving priorities over time, deconstruction of language and meaning for conceptual consistency and clarity, understanding of potential overlap and associations between outcomes, and an assessment of applicability to the range of interventions in hemodialysis. For implementation, they proposed that core outcomes must have simple, inexpensive, and validated outcome measures that could be used in clinical care (quality indicators) and trials (including pragmatic trials) and endorsement by regulatory agencies. Integrating these recommendations may foster acceptance and optimize the uptake and translation of core outcomes in hemodialysis, leading to more informative research, for better treatment and improved patient outcomes. Copyright © 2016 National Kidney Foundation, Inc. Published by Elsevier Inc. All rights reserved.

  20. Establishing Core Outcome Domains in Hemodialysis: Report of the Standardized Outcomes in Nephrology−Hemodialysis (SONG-HD) Consensus Workshop

    Science.gov (United States)

    Tong, Allison; Manns, Braden; Hemmelgarn, Brenda; Wheeler, David C.; Evangelidis, Nicole; Tugwell, Peter; Crowe, Sally; Van Biesen, Wim; Winkelmayer, Wolfgang C.; O’Donoghue, Donal; Tam-Tham, Helen; Shen, Jenny; Pinter, Jule; Larkins, Nicholas; Youssouf, Sajeda; Mandayam, Sreedhar; Ju, Angela; Craig, Jonathan C.

    2017-01-01

    Evidence-informed decision-making in clinical care and policy in nephrology is undermined by trials that selectively report a large number of heterogeneous outcomes, many of which are not patient-centered. The Standardized Outcomes in Nephrology−Hemodialysis (SONG-HD) Initiative convened an international consensus workshop on November 7, 2015, to discuss the identification and implementation of a potential core outcome set for all trials in hemodialysis. The purpose of this article is to report qualitative analyses of the workshop discussions, describing the key aspects to consider when establishing core outcomes in trials involving patients on hemodialysis. Key stakeholders including eight patients/caregivers and 47 health professionals (nephrologists, policy makers, industry, researchers) attended the workshop. Attendees suggested that identifying core outcomes required equitable stakeholder engagement to ensure relevance across patient populations; flexibility to consider evolving priorities over time; deconstruction of language and meaning for conceptual consistency and clarity; understanding of potential overlap and associations between outcomes; and an assessment of applicability to the range of interventions in hemodialysis. For implementation, they proposed that core outcomes must have simple, inexpensive and validated outcome measures that could be used in clinical care (quality ndicators) and trials (including pragmatic trials), and endorsement by regulatory agencies. Integrating these recommendations may foster acceptance and optimize the uptake and translation of core outcomes in hemodialysis, leading to more informative research, for better treatment, and improved patient outcomes. PMID:27497527

  1. The Uncommon Core

    Science.gov (United States)

    Ohler, Jason

    2013-01-01

    This author contends that the United States neglects creativity in its education system. To see this, he states, one may look at the Common Core State Standards. If one searches the English Language Arts and Literacy standards for the words "creative," "innovative," and "original"--and any associated terms, one will…

  2. Can Psychiatric Rehabilitation Be Core to CORE?

    Science.gov (United States)

    Olney, Marjorie F.; Gill, Kenneth J.

    2016-01-01

    Purpose: In this article, we seek to determine whether psychiatric rehabilitation principles and practices have been more fully incorporated into the Council on Rehabilitation Education (CORE) standards, the extent to which they are covered in four rehabilitation counseling "foundations" textbooks, and how they are reflected in the…

  3. Utah's New Mathematics Core

    Science.gov (United States)

    Utah State Office of Education, 2011

    2011-01-01

    Utah has adopted more rigorous mathematics standards known as the Utah Mathematics Core Standards. They are the foundation of the mathematics curriculum for the State of Utah. The standards include the skills and understanding students need to succeed in college and careers. They include rigorous content and application of knowledge and reflect…

  4. The Experience of Storage and Shipment for Reprocessing of HEU Nuclear Fuel Irradiated in the IRT-M Research Reactor and Pamir-630 Mobile Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sikorin, S. N.; Polazau, S. A.; Luneu, A. N.; Hrigarovich, T. K. [Joint Institute for Power and Nuclear Research–Sosny of the National Academy of Sciences of Belarus, Minsk (Belarus)

    2014-08-15

    At the end of 2010 under the Global Threat Reduction Initiative (GTRI), the Joint Institute for Power and Nuclear Research–“Sosny” (JIPNR–Sosny) of the National Academy of Sciences of the Republic of Belarus repatriated HEU spent nuclear fuel to the Russian Federation. The spent nuclear fuel was from the decommissioned Pamir-630D mobile reactor and IRT-M research reactor. The paper discusses the Pamir-630D spent nuclear fuel; experience and problems of spent nuclear fuel storage; and various aspects of the shipment including legal framework, preparation activities and shipment logistics. The conceptual project of a new research reactor for Belarus is also presented.

  5. Radiological Control Technician: Standardized technician Qualification Standard

    International Nuclear Information System (INIS)

    1992-10-01

    The Qualification Standard states and defines the knowledge and skill requirements necessary for successful completion of the Radiological Control Technician Training Program. The standard is divided into three phases: Phase I concerns RCT Academic training. There are 13 lessons associated with the core academics program and 19 lessons associated with the site academics program. The staff member should sign the appropriate blocks upon successful completion of the examination for that lesson or group of lessons. In addition, facility specific lesson plans may be added to meet the knowledge requirements in the Job Performance Measures (JPM) of the practical program. Phase II concerns RCT core/site practical (JPMs) training. There are thirteen generic tasks associated with the core practical program. Both the trainer/evaluator and student should sign the appropriate block upon successful completion of the JPM. In addition, facility specific tasks may be added or generic tasks deleted based on the results of the facility job evaluation. Phase III concerns the oral examination board successful completion of the oral examination board is documented by the signature of the chairperson of the board. Upon completion of all of the standardized technician qualification requirements, final qualification is verified by the student and the manager of the Radiological Control Department and acknowledged by signatures on the qualification standard. The completed Qualification Standard shall be maintained as an official training record

  6. Development of national standardized all-hazard disaster core competencies for acute care physicians, nurses, and EMS professionals.

    Science.gov (United States)

    Schultz, Carl H; Koenig, Kristi L; Whiteside, Mary; Murray, Rick

    2012-03-01

    The training of medical personnel to provide care for disaster victims is a priority for the physician community, the federal government, and society as a whole. Course development for such training guided by well-accepted standardized core competencies is lacking, however. This project identified a set of core competencies and performance objectives based on the knowledge, skills, and attitudes required by the specific target audience (emergency department nurses, emergency physicians, and out-of-hospital emergency medical services personnel) to ensure they can treat the injuries and illnesses experienced by victims of disasters regardless of cause. The core competencies provide a blueprint for the development or refinement of disaster training courses. This expert consensus project, supported by a grant from the Robert Wood Johnson Foundation, incorporated an all-hazard, comprehensive emergency management approach addressing every type of disaster to minimize the effect on the public's health. An instructional systems design process was used to guide the development of audience-appropriate competencies and performance objectives. Participants, representing multiple academic and provider organizations, used a modified Delphi approach to achieve consensus on recommendations. A framework of 19 content categories (domains), 19 core competencies, and more than 90 performance objectives was developed for acute medical care personnel to address the requirements of effective all-hazards disaster response. Creating disaster curricula and training based on the core competencies and performance objectives identified in this article will ensure that acute medical care personnel are prepared to treat patients and address associated ramifications/consequences during any catastrophic event. Copyright © 2012 American College of Emergency Physicians. Published by Mosby, Inc. All rights reserved.

  7. Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

    International Nuclear Information System (INIS)

    Eun, Ki Lee.; Yong, Hee Kim.; Jybe, Ho Cha.; Moon, Ghu Park.

    2000-01-01

    A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

  8. Multi-core processing and scheduling performance in CMS

    International Nuclear Information System (INIS)

    Hernández, J M; Evans, D; Foulkes, S

    2012-01-01

    Commodity hardware is going many-core. We might soon not be able to satisfy the job memory needs per core in the current single-core processing model in High Energy Physics. In addition, an ever increasing number of independent and incoherent jobs running on the same physical hardware not sharing resources might significantly affect processing performance. It will be essential to effectively utilize the multi-core architecture. CMS has incorporated support for multi-core processing in the event processing framework and the workload management system. Multi-core processing jobs share common data in memory, such us the code libraries, detector geometry and conditions data, resulting in a much lower memory usage than standard single-core independent jobs. Exploiting this new processing model requires a new model in computing resource allocation, departing from the standard single-core allocation for a job. The experiment job management system needs to have control over a larger quantum of resource since multi-core aware jobs require the scheduling of multiples cores simultaneously. CMS is exploring the approach of using whole nodes as unit in the workload management system where all cores of a node are allocated to a multi-core job. Whole-node scheduling allows for optimization of the data/workflow management (e.g. I/O caching, local merging) but efficient utilization of all scheduled cores is challenging. Dedicated whole-node queues have been setup at all Tier-1 centers for exploring multi-core processing workflows in CMS. We present the evaluation of the performance scheduling and executing multi-core workflows in whole-node queues compared to the standard single-core processing workflows.

  9. In-core Instrument Subcritical Verification (INCISV) - Core Design Verification Method - 358

    International Nuclear Information System (INIS)

    Prible, M.C.; Heibel, M.D.; Conner, S.L.; Sebastiani, P.J.; Kistler, D.P.

    2010-01-01

    According to the standard on reload startup physics testing, ANSI/ANS 19.6.1, a plant must verify that the constructed core behaves sufficiently close to the designed core to confirm that the various safety analyses bound the actual behavior of the plant. A large portion of this verification must occur before the reactor operates at power. The INCISV Core Design Verification Method uses the unique characteristics of a Westinghouse Electric Company fixed in-core self powered detector design to perform core design verification after a core reload before power operation. A Vanadium self powered detector that spans the length of the active fuel region is capable of confirming the required core characteristics prior to power ascension; reactivity balance, shutdown margin, temperature coefficient and power distribution. Using a detector element that spans the length of the active fuel region inside the core provides a signal of total integrated flux. Measuring the integrated flux distributions and changes at various rodded conditions and plant temperatures, and comparing them to predicted flux levels, validates all core necessary core design characteristics. INCISV eliminates the dependence on various corrections and assumptions between the ex-core detectors and the core for traditional physics testing programs. This program also eliminates the need for special rod maneuvers which are infrequently performed by plant operators during typical core design verification testing and allows for safer startup activities. (authors)

  10. Prompt Neutron Lifetime for the NBSR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2012-06-24

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  11. SedCT: MATLAB™ tools for standardized and quantitative processing of sediment core computed tomography (CT) data collected using a medical CT scanner

    Science.gov (United States)

    Reilly, B. T.; Stoner, J. S.; Wiest, J.

    2017-08-01

    Computed tomography (CT) of sediment cores allows for high-resolution images, three-dimensional volumes, and down core profiles. These quantitative data are generated through the attenuation of X-rays, which are sensitive to sediment density and atomic number, and are stored in pixels as relative gray scale values or Hounsfield units (HU). We present a suite of MATLAB™ tools specifically designed for routine sediment core analysis as a means to standardize and better quantify the products of CT data collected on medical CT scanners. SedCT uses a graphical interface to process Digital Imaging and Communications in Medicine (DICOM) files, stitch overlapping scanned intervals, and create down core HU profiles in a manner robust to normal coring imperfections. Utilizing a random sampling technique, SedCT reduces data size and allows for quick processing on typical laptop computers. SedCTimage uses a graphical interface to create quality tiff files of CT slices that are scaled to a user-defined HU range, preserving the quantitative nature of CT images and easily allowing for comparison between sediment cores with different HU means and variance. These tools are presented along with examples from lacustrine and marine sediment cores to highlight the robustness and quantitative nature of this method.

  12. Design studies of back up cores for the experimental multi-purpose VHTR, (1)

    International Nuclear Information System (INIS)

    Yasuno, Takehiko; Miyamoto, Yoshiaki; Mitake, Susumu

    1982-09-01

    For the Experimental Multi-Purpose Very High Temperature Reactor, design studies have been made of two backup cores loaded with new type fuel elements. The purpose is to improve core operational characteristics of the standard design core (Mark-III core) consisting of pin-in-block type fuel element having externally cooled hollow fuel rods. The first backup core (semi-pin fuel core) is composed of fuel elements with internally cooled fuel pins, and the second core (multihole fuel core) is composed of multihole fuel elements, which can be adopted for the experimental VHTR as the substitution of the standard Mark-III fuel element. Either of the cores has 73 fuel columns and 4 m height. The arrangement of active core and reactor internal structure is same as that in the standard design core. These backup cores meet almost all design requirements of the VHTR and increase the margins for some important design items in comparison with the standard core (Mark-III core). This report describes the overall characteristics of nuclear, thermal-hydraulic, fuel and safety, and structural consideration for these cores. (author)

  13. Common Core State Standards and Teacher Effectiveness. Q&A with Ross Wiener, Ph.D. REL Mid-Atlantic Teacher Effectiveness Webinar Series

    Science.gov (United States)

    Regional Educational Laboratory Mid-Atlantic, 2013

    2013-01-01

    In this REL Mid-Atlantic webinar, Dr. Ross Wiener, Vice President and Executive Director of the Education and Society Program, Aspen Institute, discussed strategies for integrating the Common Core State Standards (CCSS) into teacher effectiveness systems, including ways in which the CCSS can support professional growth and inform teacher…

  14. Multi-core processing and scheduling performance in CMS

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    Commodity hardware is going many-core. We might soon not be able to satisfy the job memory needs per core in the current single-core processing model in High Energy Physics. In addition, an ever increasing number of independent and incoherent jobs running on the same physical hardware not sharing resources might significantly affect processing performance. It will be essential to effectively utilize the multi-core architecture. CMS has incorporated support for multi-core processing in the event processing framework and the workload management system. Multi-core processing jobs share common data in memory, such us the code libraries, detector geometry and conditions data, resulting in a much lower memory usage than standard single-core independent jobs. Exploiting this new processing model requires a new model in computing resource allocation, departing from the standard single-core allocation for a job. The experiment job management system needs to have control over a larger quantum of resource since multi-...

  15. State Standard-Setting Processes in Brief. State Academic Standards: Standard-Setting Processes

    Science.gov (United States)

    Thomsen, Jennifer

    2014-01-01

    Concerns about academic standards, whether created by states from scratch or adopted by states under the Common Core State Standards (CCSS) banner, have drawn widespread media attention and are at the top of many state policymakers' priority lists. Recently, a number of legislatures have required additional steps, such as waiting periods for…

  16. Competitive Debate as Competency-Based Learning: Civic Engagement and Next-Generation Assessment in the Era of the Common Core Learning Standards

    Science.gov (United States)

    McIntosh, Jonathan; Milam, Myra

    2016-01-01

    As the adoption and execution of the Common Core State Standards (CCSS) have steadily increased, the debate community is presented with an opportunity to be more forward thinking and sustainable through the translation to curriculum planning and next-generation assessment as a movement towards Performance-Based Assessments. This paper focuses on…

  17. A neutronic feasibility study for LEU conversion of the WWR-M reactor at Gatchina

    International Nuclear Information System (INIS)

    Petrov, Yu. V.; Erykalov, A.N.; Onegin, M.S.

    2000-01-01

    In this report we present the results of computations of the full scale reactor core with HEU (90%), MEU (36%) and LEU (19.75%) fuel. The reactor computer model for the MCU RFFI Monte Carlo code includes all peculiarities of the core. Calculations show that a uranium density of 3.3gU/cm 3 of MEU (36%) fuel and 8/25gU/cm 3 of LEU (19.75%) in WWR-M5 fuel assembly (FA) geometry is required to match the fuel cycle length of the HEU (90%) case with the same end of cycle (EOEC) excess reactivity. For the equilibrium fuel cycle the fuel burnup and poisoning, the fast and thermal neutron fluxes, the reactivity worth of control rods were calculated for the reference case with HEU (90%) FA and for the MEU and LEU FA. The relative accuracy of this neutronic feasibility study of fuel enrichment reduction of the WWR-M reactor in Gatchina is sufficient to start the fabrication feasibility study of MEU (36%) WWR-M5 fuel assemblies. At the present stage of technology it seems hardly possible to manufacture LEU (19.75%) fuel elements in WWR-M5 geometry due to too high uranium density. Only a future R and D can solve the problem. (author)

  18. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

  19. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  20. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  1. A "Common" Vision of Instruction? An Analysis of English/Language Arts Professional Development Materials Related to the Common Core State Standards

    Science.gov (United States)

    Hodge, Emily; Benko, Susanna L.

    2014-01-01

    The purpose of this article is to describe the stances put forward by a selection of professional development resources interpreting the Common Core State Standards for English Language Arts (ELA) teachers, and to analyse where these resources stand in relation to research in ELA. Specifically, we analyse resources written by English educators…

  2. Isotope Enrichment Detection by Laser Ablation - Laser Absorption Spectrometry: Automated Environmental Sampling and Laser-Based Analysis for HEU Detection

    International Nuclear Information System (INIS)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-01-01

    The global expansion of nuclear power, and consequently the uranium enrichment industry, requires the development of new safeguards technology to mitigate proliferation risks. Current enrichment monitoring instruments exist that provide only yes/no detection of highly enriched uranium (HEU) production. More accurate accountancy measurements are typically restricted to gamma-ray and weight measurements taken in cylinder storage yards. Analysis of environmental and cylinder content samples have much higher effectiveness, but this approach requires onsite sampling, shipping, and time-consuming laboratory analysis and reporting. Given that large modern gaseous centrifuge enrichment plants (GCEPs) can quickly produce a significant quantity (SQ ) of HEU, these limitations in verification suggest the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory (PNNL) is developing an unattended safeguards instrument concept, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely analysis of enrichment levels within low enriched uranium facilities. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy to characterize the uranium isotopic ratio through subtle differences in atomic absorption wavelengths. Environmental sampling (ES) media from an integrated aerosol collector is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes material from a 10 to 20-(micro)m diameter spot of the surface of the sampling media. The plume of ejected material begins as high-temperature plasma that yields ions and atoms, as well as molecules and molecular ions. We concentrate on the plume of atomic vapor that remains after the plasma has expanded and then cooled by the surrounding cover gas. Tunable diode lasers are directed through this plume and each isotope is detected by monitoring absorbance

  3. Math starters 5- to 10-minute activities aligned with the common core math standards, grades 6-12

    CERN Document Server

    Muschla, Judith A; Muschla, Erin

    2013-01-01

    A revised edition of the bestselling activities guide for math teachers Now updated with new math activities for computers and mobile devices-and now organized by the Common Core State Standards-this book includes more than 650 ready-to-use math starter activities that get kids quickly focused and working as soon as they enter the classroom. Ideally suited for any math curriculum, these high-interest problems spark involvement in the day's lesson, help students build skills, and allow teachers to handle daily management tasks without wasting valuable instructional time. A newly updated edit

  4. .net core application lifecycle on Openshift

    CERN Multimedia

    CERN. Geneva

    2017-01-01

    # .net core application lifecycle on Openshift I will show an example of a lifecycle of an OpenShift application with an emphasis on the continuous integration and deployment. The application compatible with [.net Standard](https://docs.microsoft.com/en-us/dotnet/standard/net-standard) can be easily deployed on OpenShift using [Source2Image](https://docs.openshift.com/enterprise/3.0/architecture/core_concepts/builds_and_image_streams.html#source-build) functionality, which doesn't require developers to maintain docker images of the application. I will also present how to efficiently integrate this feature into GitLab pipelines with an automated deployment of the "review" environment, as one its parts.

  5. Common Core: Teaching Optimum Topic Exploration (TOTE)

    Science.gov (United States)

    Karge, Belinda Dunnick; Moore, Roxane Kushner

    2015-01-01

    The Common Core has become a household term and yet many educators do not understand what it means. This article explains the historical perspectives of the Common Core and gives guidance to teachers in application of Teaching Optimum Topic Exploration (TOTE) necessary for full implementation of the Common Core State Standards. An effective…

  6. Thermal hydraulic design of a hydride-fueled inverted PWR core

    International Nuclear Information System (INIS)

    Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

    2009-01-01

    An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

  7. Integration of Biosafety into Core Facility Management

    OpenAIRE

    Fontes, Benjamin

    2013-01-01

    This presentation will discuss the implementation of biosafety policies for small, medium and large core laboratories with primary shared objectives of ensuring the control of biohazards to protect core facility operators and assure conformity with applicable state and federal policies, standards and guidelines. Of paramount importance is the educational process to inform core laboratories of biosafety principles and policies and to illustrate the technology and process pathways of the core l...

  8. Simplifying the ELA Common Core; Demystifying Curriculum

    Science.gov (United States)

    Schmoker, Mike; Jago, Carol

    2013-01-01

    The English Language Arts (ELA) Common Core State Standards ([CCSS], 2010) could have a transformational effect on American education. Though the process seems daunting, one can begin immediately integrating the essence of the ELA Common Core in every subject area. This article shows how one could implement the Common Core and create coherent,…

  9. Performance evaluation of converted and upgraded PARR-1

    International Nuclear Information System (INIS)

    Pervez, Showket; Iqbal, Masood

    1998-01-01

    Pakistan Research Reactor-1 (PARR-1), a swimming pool MTR type research reactor which attained full power of 5 MW in June, 1996, with 93% high enriched uranium (HEU) fuel was converted to ,20% Low Enriched Uranium (LEU) fuel in october, 1991. The reactor power was also upgraded from 5 MW to 9 MW and then to 10 MW. Different critical and full power operational core configurations were assembled with the new fuel. The final equilibrium core was assembled with 27 standard fuel elements and five control fuel elements having a central flux trap facility for high neutron flux. detailed neutronics and thermal-hydraulic design calculations were made for the core conversion programme. After achieving the initial criticality several critical and power experiments were performed on the new core for the verification of design data and to determine the nuclear performance of the reactor. a comparison of the measured and the calculated results was also made. the results of the characteristics tests indicate that the performance of the new reactor is within the design limits. In flux trap thermal neutron flux is about 2x10 14 n.cm - 2. s - 1 which is five times higher than the average neutron flux of the core, seven standard and two control fuel elements have achieved designed burnup of 35%. Their physical inspection predicts excellent condition. (author)

  10. Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

    International Nuclear Information System (INIS)

    Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

    2004-01-01

    The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

  11. Birthweight, HIV exposure and infant feeding as predictors of malnutrition in Botswanan infants.

    Science.gov (United States)

    Chalashika, P; Essex, C; Mellor, D; Swift, J A; Langley-Evans, S

    2017-12-01

    A better understanding of the nutritional status of infants who are HIV-Exposed-Uninfected (HEU) and HIV-Unexposed-Uninfected (HUU) during their first 1000 days is key to improving population health, particularly in sub-Saharan Africa. A cross-sectional study compared the nutritional status, feeding practices and determinants of nutritional status of HEU and HUU infants residing in representative selected districts in Botswana during their first 1000 days of life. Four hundred and thirteen infants (37.3% HIV-exposed), aged 6-24 months, attending routine child health clinics, were recruited. Anthropometric, 24-h dietary intake and socio-demographic data was collected. Anthropometric Z-scores were calculated using 2006 World Health Organization growth standards. Modelling of the determinants of malnutrition was undertaken using logistic regression. Overall, the prevalences of stunting, wasting and being underweight were 10.4%, 11.9% and 10.2%, respectively. HEU infants were more likely to be underweight (15.6% versus 6.9%), (P economic status. HEU infants aged 6-24 months had worse nutritional status compared to HUU infants. Low birthweight was the main predictor of undernutrition in this population. Optimisation of infant nutritional status should focus on improving birthweight. In addition, specific interventions should target HEU infants aiming to eliminate growth disparity between HEU and HUU infants. © 2017 The British Dietetic Association Ltd.

  12. Russian RERTR program as a part of Joint US DOE-RF MINATOM collaboration on elimination of the threat connected to the use of HEU in research reactors

    International Nuclear Information System (INIS)

    Arkhangelsky, N.

    2002-01-01

    The Russian RERTR Program started at the end of 70's, the final goal of the program is to eliminate supplies of HEU in fuel elements and assemblies for foreign research reactors that were designed according to Russian projects. Basic directions of the work include: completion of the development of the fuel elements and assemblies on a basis of uranium dioxide; development of the fuel on a basis of U-Mo alloy; and development of pin type fuel elements. Fuel assemblies of WWR-M2 type with LEU were developed and qualified for using in foreign research reactors that use such type of fuel assemblies. These assemblies are ready for the supplying several operating foreign research reactors. There are more than 20 sites in Eastern European countries, former Soviet republics and another countries that have big amount of Russian origin HEU in fresh and spent fuel. The problem of the shipment of SNF from sites of research reactors is also very important for domestic Russian research reactors. More than ten years from its beginning the Russian RERTR program developed practically independently from the international RERTR program and only at the begin of 90's the Russian specialists started to contact with foreign scientists and the exchange of the scientific information has become more intensive. In September 1994, representatives of Minatom and DOE signed a protocol of intent to reduce an enrichment of uranium in research reactors. The main aspects of collaboration involve: Several domestic Russian research reactors such as WWR-M, IR-8 and others were investigated from the point of view of possibility of reducing of enrichment; financial support of the program from US DOE which is insufficient. The important part of international collaboration is the import of Russian origin spent and fresh fuel of research reactors to Russia. In August 2002 an impressive result of the Russian-American collaboration with support of IAEA and with the help and assistance of Yugoslavian side was

  13. A new standard for core training in radiation safety

    International Nuclear Information System (INIS)

    Trinoskey, P.A.

    1997-02-01

    A new American National Standard for radiation worker training was recently developed. The standard emphasizes performance-based training and establishing a training program rather than simply prescribing objectives. The standard also addresses basic criteria, including instructor qualifications. The standard is based on input from a wide array of regulatory agencies, universities, national laboratories, and nuclear power entities. This paper presents an overview of the new standard and the philosophy behind it. The target audience includes radiation workers, management and supervisory personnel, contractors, students, emergency personnel, and visitors

  14. Ultrabroadband polarization splitter based on three-core photonic crystal fiber with a modulation core.

    Science.gov (United States)

    Zhao, Tongtong; Lou, Shuqin; Wang, Xin; Zhou, Min; Lian, Zhenggang

    2016-08-10

    We design an ultrabroadband polarization splitter based on three-core photonic crystal fiber (PCF). A modulation core and two fluorine-doped cores are introduced to achieve an ultrawide bandwidth. The properties of three-core PCF are modeled by using the full-vector finite element method along with the full-vector beam propagation method. Numerical results demonstrate that an ultrabroadband splitter with 320 nm bandwidth with an extinction ratio as low as -20  dB can be achieved by using 52.8 mm long three-core PCF. This splitter also has high compatibility with standard single-mode fibers as the input and output ports due to low splicing loss of 0.02 dB. All the air holes in the proposed structure are circular holes and arranged in a triangular lattice that makes it easy to fabricate.

  15. Neutronic calculation of safety parameters for the RP-0 and RP-10 nuclear reactors

    OpenAIRE

    Lázaro, Gerardo; Deen, James R.; Woodruff, William L.

    2002-01-01

    Theoretical safety calculations were done with proved codes utilized by the staff of the RERTR program in the HEU to LEU core conversions. The studies were designed to evaluate the reactivity coefficients and kinetics parameters of the reactor involved in the evolution of peak power transients by reactivity insertion accidents. It was done to show the trend of these reactivity coefficients as a function of the core size and fuel depletion for RP10 cores. It was useful to get a better underst...

  16. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  17. Common Core State Standards in the Middle Grades: What's New in the Geometry Domain and How Can Teachers Support Student Learning?

    Science.gov (United States)

    Teuscher, Dawn; Tran, Dung; Reys, Barbara J.

    2015-01-01

    The Common Core State Standards for Mathematics (CCSSM) is a primary focus of attention for many stakeholders' (e.g., teachers, district mathematics leaders, and curriculum developers) intent on improving mathematics education. This article reports on specific content shifts related to the geometry domain in the middle grades (6-8)…

  18. Comparisons of significant parameters for a standard 20% enriched and FLIP 70% enriched TRIGA core

    International Nuclear Information System (INIS)

    Ringle, John C.; Anderson, Terrance V.; Johnson, Arthur G.

    1978-01-01

    A comparison is made between the 20% and 70% enriched cores. The initial start-up data for both cores show the FLIP needs ∼3.8 times the 235 U mass as the 20% core just to go critical. Operational configurations for both cores indicate a need for ∼33% additional fuel above initial critical for adequate maneuvering excess. The fuel element worths are higher in the central core locations for the 20% elements while the peripheral element worths are about the same (with some thermal flux peaking in the FLIP perheral elements). Pulsing comparisons of the two cores show significant differences in reactivity insertions and power peaks. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

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

  20. Use of highly enriched uranium at the FRM-II

    Energy Technology Data Exchange (ETDEWEB)

    Boening, K. [Forschungs-Neutronenquelle FRM-II, Technische Universitaet Muenchen, D-85747 Garching bei Muenchen (Germany)

    2002-07-01

    The new FRM-II research reactor in Munich, Germany, provides a high flux of thermal neutrons outside of the core at only 20 MW power. This is achieved by using a single compact, cylindrical fuel element with highly enriched uranium (HEU) which is cooled by light water and placed in the center of a large heavy water tank. The paper outlines the arguments which have led to this core concept and summarizes its performance. It also reports on alternative studies which have been performed for the case of low enriched uranium (LEU) and compares the data of the two concepts, with the conclusion that the FRM-II cannot be converted to LEU. A concept using medium enriched uranium (MEU) is described as well as plans to develop such a fuel element in the future. Finally, it is argued that the use of HEU fuel elements at the FRM-II does not - realistically -involve any risk of proliferation. (author)

  1. Upper-bound fission product release assessment for large break LOCA in CANFLEX bundle reactor core

    International Nuclear Information System (INIS)

    Oh, Duk Ju; Lee, Kang Moon

    1996-07-01

    Quarter-core gap inventory assessment for CANDU-6 reactor core loaded with CANFLEX fuel bundles has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The quarter-core gap inventory for the CANFLEX bundle core is 5 - 10 times lower than that for the standard bundle core, depending on the half-life of the isotope. The lower gap inventory of the CANFLEX bundle core is attributed to the lower linear power of the CANFLEX bundle compared with the standard bundle. However, the whole core total inventories for both the CANFLEX and standard bundle cores are nearly the same. The 6 - 8 times lower upper-bound fission product releases of the CANFLEX bundle core for large break LOCA than those of the standard bundle core imply that the loading of 24 natural uranium CANFLEX bundles would improve the predicted consequences of the postulated accident described in the Wolsung 2 safety report. 2 tabs., 6 figs., 3 refs. (Author)

  2. The Influences of Middle School Mathematics Teachers' Practical Rationality on Instructional Decision Making Regarding the Common Core State Standards for Mathematical Practices

    Science.gov (United States)

    Sobolewski-McMahon, Lauren M.

    2017-01-01

    The purpose of this study was to examine the influences of various facets of middle school mathematics teachers' practical rationality on their instructional decision making as they plan to enact the Common Core State Standards for Mathematical Practice, CCSS-MP1 (perseverance in problem solving) and CCSS-MP3 (communicating and critiquing). The…

  3. Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

    1998-10-01

    Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

  4. Mathematical Communication in State Standards before the Common Core

    Science.gov (United States)

    Kosko, Karl Wesley; Gao, Yang

    2017-01-01

    Mathematical communication has been an important feature of standards documents since National Council of Teachers of Mathematics' (NCTM) (1989) "Curriculum and Evaluation Standards." Such an emphasis has influenced content standards of states from then to present. This study examined how effective the prevalence of various forms of…

  5. On-line core monitoring system based on buckling corrected modified one group model

    International Nuclear Information System (INIS)

    Freire, Fernando S.

    2011-01-01

    Nuclear power reactors require core monitoring during plant operation. To provide safe, clean and reliable core continuously evaluate core conditions. Currently, the reactor core monitoring process is carried out by nuclear code systems that together with data from plant instrumentation, such as, thermocouples, ex-core detectors and fixed or moveable In-core detectors, can easily predict and monitor a variety of plant conditions. Typically, the standard nodal methods can be found on the heart of such nuclear monitoring code systems. However, standard nodal methods require large computer running times when compared with standards course-mesh finite difference schemes. Unfortunately, classic finite-difference models require a fine mesh reactor core representation. To override this unlikely model characteristic we can usually use the classic modified one group model to take some account for the main core neutronic behavior. In this model a course-mesh core representation can be easily evaluated with a crude treatment of thermal neutrons leakage. In this work, an improvement made on classic modified one group model based on a buckling thermal correction was used to obtain a fast, accurate and reliable core monitoring system methodology for future applications, providing a powerful tool for core monitoring process. (author)

  6. Diffusion calculation's for the SLOWPOKE-2 reactor using DONJON

    International Nuclear Information System (INIS)

    Noceir, S.; El Hajjaji, O.; Varin, E.

    1997-01-01

    The SLOWPOKE reactor at Ecole Polytechnique will be refueled with a Low Enriched Uranium (LEU) fuel in place of a High Enriched Uranium (HEU) fuel used until now. The purpose of this study is to provide various models, using the reactor physics chain of codes DRAGON/DONJON, in order to predict the behavior of the new LEU Slowpoke. In particle, we will present some numerical results concerning the separate temperature effects of the main components of the core, the effect of a partial void appearing near the fuel pins and the axial and radial flux distributions. Finally the difference between the present HEU and the future LEU fuel power will be given. (author)

  7. Revisiting Traveling Books: Early Literacy, Social Studies, and the Common Core

    Science.gov (United States)

    Swain, Holly Hilboldt; Coleman, Julianne

    2015-01-01

    With the development and institution of the Common Core Standards, teachers must be prepared to integrate content areas such as social studies within the language arts curriculum. Teachers following the suggestions of the Common Core Standards should develop practical and meaningful strategies within their classrooms that encourage and support…

  8. Back up core designs for the experimental multi-purpose VHTR

    International Nuclear Information System (INIS)

    Aochi, Tetsuo; Yasuno, Takehiko; Miyamoto, Yoshiaki; Shindo, Ryuichi; Ikushima, Takeshi

    1979-02-01

    For the Experimental Multi-Purpose Very High Temperature Reactor (thermal power 50 MW and reactor outlet helium temperature 1000 0 C), design studies have been made of two backup cores loaded with new-type fuel elements. The purpose is to improve core operational characteristics, especially in thermohydraulics, of the reference design core consisting of pin-in-block type fuel elements having externally cooled hollow fuel rods. In this report are described the design principles and the analyses made of nuclear, thermal and hydraulic, fuel, and safety performances to determine the backup fuel and core design parameters. The first backup core (SP fuel core) is composed of fuel elements with internally cooled fuel rods (semi-pin), 36 rods in each standard element and 18 rods in each control element. The second backup core (MH fuel core) is composed of multihole fuel elements. 102 fuel and 54 coolant holes in each standard element and 30 fuel and 18 coolant holes in each control element. Either of the cores has 73 fuel columns 4 m high; the arrangement of active core and reactor internal structures is the same as that in the reference design. The backup cores meet nearly all design requirements of the VHTR, permitting the rated power operation with coolant Reynolds number of over 10,000 in the SP core and over 6,000 in the MH core. (author)

  9. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  10. [caCORE: core architecture of bioinformation on cancer research in America].

    Science.gov (United States)

    Gao, Qin; Zhang, Yan-lei; Xie, Zhi-yun; Zhang, Qi-peng; Hu, Zhang-zhi

    2006-04-18

    A critical factor in the advancement of biomedical research is the ease with which data can be integrated, redistributed and analyzed both within and across domains. This paper summarizes the Biomedical Information Core Infrastructure built by National Cancer Institute Center for Bioinformatics in America (NCICB). The main product from the Core Infrastructure is caCORE--cancer Common Ontologic Reference Environment, which is the infrastructure backbone supporting data management and application development at NCICB. The paper explains the structure and function of caCORE: (1) Enterprise Vocabulary Services (EVS). They provide controlled vocabulary, dictionary and thesaurus services, and EVS produces the NCI Thesaurus and the NCI Metathesaurus; (2) The Cancer Data Standards Repository (caDSR). It provides a metadata registry for common data elements. (3) Cancer Bioinformatics Infrastructure Objects (caBIO). They provide Java, Simple Object Access Protocol and HTTP-XML application programming interfaces. The vision for caCORE is to provide a common data management framework that will support the consistency, clarity, and comparability of biomedical research data and information. In addition to providing facilities for data management and redistribution, caCORE helps solve problems of data integration. All NCICB-developed caCORE components are distributed under open-source licenses that support unrestricted usage by both non-profit and commercial entities, and caCORE has laid the foundation for a number of scientific and clinical applications. Based on it, the paper expounds caCORE-base applications simply in several NCI projects, of which one is CMAP (Cancer Molecular Analysis Project), and the other is caBIG (Cancer Biomedical Informatics Grid). In the end, the paper also gives good prospects of caCORE, and while caCORE was born out of the needs of the cancer research community, it is intended to serve as a general resource. Cancer research has historically

  11. From time-based to competency-based standards: core transitional competencies in plastic surgery.

    Science.gov (United States)

    Lutz, Kristina; Yazdani, Arjang; Ross, Douglas

    2015-01-01

    Competency-based medical education is becoming increasingly prevalent and is likely to be mandated by the Royal College in the near future. The objective of this study was to define the core technical competencies that should be possessed by plastic surgery residents as they transition into their senior (presently postgraduate year 3) years of training. A list of potential core competencies was generated using a modified Delphi method that included the investigators and 6 experienced, academic plastic surgeons from across Canada and the United States. Generated items were divided into 7 domains: basic surgical skills, anesthesia, hand surgery, cutaneous surgery, esthetic surgery, breast surgery, and craniofacial surgery. Members of the Delphi group were asked to rank particular skills on a 4-point scale with anchored descriptors. Item reduction resulted in a survey consisting of 48 skills grouped into the aforementioned domains. This self-administered survey was distributed to all Canadian program directors (n = 11) via e-mail for validation and further item reduction. The response rate was 100% (11/11). Using the average rankings of program directors, 26 "core" skills were identified. There was agreement of core skills across all domains except for breast surgery and esthetic surgery. Of them, 7 skills were determined to be above the level of a trainee at this stage; a further 15 skills were agreed to be important, but not core, competencies. Overall, 26 competencies have been identified as "core" for plastic surgery residents to possess as they begin their senior, on-service years. The nature of these skills makes them suitable for teaching in a formal, simulated environment, which would ensure that all plastic surgery trainees are competent in these tasks as they transition to their senior years of residency. Copyright © 2014 Association of Program Directors in Surgery. Published by Elsevier Inc. All rights reserved.

  12. Search for sterile neutrinos with IceCube DeepCore

    Energy Technology Data Exchange (ETDEWEB)

    Terliuk, Andrii [DESY, Platanenallee 6, 15738 Zeuthen (Germany); Collaboration: IceCube-Collaboration

    2016-07-01

    The DeepCore detector is a sub-array of the IceCube Neutrino Observatory that lowers the energy threshold for neutrino detection down to approximately 10 GeV. DeepCore is used for a variety of studies including atmospheric neutrino oscillations. The standard three-neutrino oscillation paradigm is tested using the DeepCore detector by searching for an additional light, sterile neutrino with a mass on the order of 1 eV. Sterile neutrinos do not interact with the ordinary matter, however they can be mixed with the three active neutrino states. Such mixture changes the picture of standard neutrino oscillations for atmospheric neutrinos with energies below 100 GeV. The capabilities of DeepCore detector to measure such sterile neutrino mixing will be presented in this talk.

  13. Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium in Oxides within DOE-STD-3013-2000 Containers

    International Nuclear Information System (INIS)

    Mount, M E; O'Connell, W J

    2005-01-01

    Lawrence Livermore National Laboratory (LLNL) uses the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. In June 2002, at the 43rd Annual Meeting of the Institute of Nuclear Material Management, LLNL reported on an extensive effort to calibrate this shuffler, based on standards measurements and extensive simulations, for HEU oxides and mixed U-Pu oxides in thin-walled primary and secondary containers. In August 2002, LLNL began to also use DOE-STD-3013-2000 containers for HEU oxide and mixed U-Pu oxide. These DOE-STD-3013-2000 containers are comprised of a stainless steel convenience can enclosed in welded stainless steel primary and secondary containers. Compared to the double thin-walled containers, the DOE-STD-3013-2000 containers have substantially thicker walls, and the density of materials in these containers was found to extend over a greater range (1.35 g/cm 3 to 4.62 g/cm 3 ) than foreseen for the double thin-walled containers. Further, the DOE-STD-3013-2000 Standard allows for oxides containing at least 30 wt% Pu plus U whereas the calibration algorithms for thin-walled containers were derived for virtually pure HEU or mixed U-Pu oxides. An initial series of Monte Carlo simulations of the PAN shuffler response to given quantities of HEU oxide and mixed U-Pu oxide in DOE-STD-3013-2000 containers was generated and compared with the response predicted by the calibration algorithms for thin-walled containers. Results showed a decrease on the order of 10% in the count rate, and hence a decrease in the calculated U mass for measured unknowns, with some varying trends versus U mass. Therefore a decision was made to develop a calibration algorithm for the PAN shuffler unique to the DOE-STD-3013-2000 container. This paper describes that effort and selected unknown item measurement results

  14. Primary Sources. Update: Teachers' Views on Common Core State Standards

    Science.gov (United States)

    Scholastic Inc. and the Bill & Melinda Gates Foundation, 2014

    2014-01-01

    Scholastic and the Bill & Melinda Gates Foundation fielded the third edition of the "Primary Sources" survey of America's teachers in July 2013 (see ED562664). Twenty thousand pre-K through grade 12 public school teachers responded, sharing their perspectives on issues important to their profession, including the Common Core State…

  15. Fire resistance of extruded hollow-core slabs

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl; Sørensen, Lars Schiøtt; Giuliani, Luisa

    2017-01-01

    to the structural codes with data derived from a standard fire test and from a thorough examination of the comprehensive test documentation available on fire exposed hollow-core slabs. Findings – Mechanisms for loss of load-bearing capacity are clarified, and evidence of the fire resistance is found. Originality......Purpose – Prefabricated extruded hollow-core slabs are preferred building components for floor structures in several countries. It is therefore important to be able to document the fire resistance of these slabs proving fulfilment of standard fire resistance requirements of 60 and 120 min found...... in most national building regulations. The paper aims to present a detailed analysis of the mechanisms responsible for the loss of loadbearing capacity of hollow-core slabs when exposed to fire. Design/methodology/approach – Furthermore, it compares theoretica calculation and assessment according...

  16. Characteristics of core sampling from crumbing Paleozoic rock

    Energy Technology Data Exchange (ETDEWEB)

    Barabashkin, I I; Edelman, Y A; Filippov, V N; Lychev, V N

    1981-01-01

    The results of analysis of core sampling using standard core sampling tools with small and medium inside diameter are cited. It is demonstrated that when using these tools loss of core in Paleozoic deposits promising with regard to oil and gas content does not exceed 25 - 30%. The use of a new core sampling tool with a large inside diameter which includes drill bits of different types and a core lifter ''Krembriy'' SKU-172/100 made it possible to increase core removal approximately 52%. A representative core from a highly crumbling and vesicular rock belinging to groups III - IV in terms of difficulty of core sampling was obtained first. A description of a new core sampling tool is given. The characteristics of the technology of its use which promote preservation of the core are cited. Means of continued improvement of this tool are noted.

  17. Common Core Implementation Decisions Made by Principals in Elementary Schools

    Science.gov (United States)

    Norman, Alexis Cienfuegos

    2016-01-01

    The purpose of this study was to understand the decisions elementary principals have made during the Common Core State Standards reform. Specifically, (a) what decisions principals have made to support Common Core implementation, (b) what strategies elementary principals have employed to communicate with stakeholders about Common Core State…

  18. Standard deviation of local tallies in global Monte Carlo calculation of nuclear reactor core

    International Nuclear Information System (INIS)

    Ueki, Taro

    2010-01-01

    Time series methodology has been studied to assess the feasibility of statistical error estimation in the continuous space and energy Monte Carlo calculation of the three-dimensional whole reactor core. The noise propagation was examined and the fluctuation of track length tallies for local fission rate and power has been formally shown to be represented by the autoregressive moving average process of orders p and p-1 [ARMA(p,p-1)], where p is an integer larger than or equal to two. Therefore, ARMA(p,p-1) fitting was applied to the real standard deviation estimation of the power of fuel assemblies at particular heights. Numerical results indicate that straightforward ARMA(3,2) fitting is promising, but a stability issue must be resolved toward the incorporation in the distributed version of production Monte Carlo codes. The same numerical results reveal that the average performance of ARMA(3,2) fitting is equivalent to that of the batch method with a batch size larger than 100 and smaller than 200 cycles for a 1,100 MWe pressurized water reactor. (author)

  19. Installation of JMTR core management system

    International Nuclear Information System (INIS)

    Imaizumi, Tomomi; Ide, Hiroshi; Naka, Michihiro; Komukai, Bunsaku; Nagao, Yoshiharu

    2013-01-01

    In order to carry out the core management after the reoperation of JMTR quickly and accurately, the authors took up the Standard Reactor Analysis Code (SRAC) system and core management support programs that are operating in a general-purpose large computer and transferred them to PC (OS: Linux), and newly established a JMTR core management system. As for the core analysis, this measure enabled an increase in the processing speed from the check of core arrangement to the result display of nuclear restriction values to about 60 times, compared with the conventional method. It was confirmed that the differences of calculation results originated from the difference of internal display of computers, associated with the transfer of each analysis code from GS21-400 system to PC-Linux, were within practically allowable level. In the future, this system will be applied to the core analysis of JMTR, as well as to the preparation of operation plans. (A.O.)

  20. The Darwin Core extension for genebanks opens up new opportunities for sharing genebank datasets

    Directory of Open Access Journals (Sweden)

    Dag Terje Filip Endresen

    2012-07-01

    Full Text Available Darwin Core (DwC defines a standard set of terms to describe the primary biodiversity data. Primary biodiversity data are data records derived from direct observation of species occurrences in nature or describing specimens in biological collections. The Darwin Core terms can be seen as an extension to the standard Dublin Core metadata terms. The new Darwin Core extension for genebanks declares the additional terms required for describing genebank datasets, and is based on established standards from the plant genetic resources community. The Global Biodiversity Information Facility (GBIF provides an information infrastructure for biodiversity data including a suite of software tools for data publishing, distributed data access, and the capture of biodiversity data. The Darwin Core extension for genebanks is a key component that provides access for the genebanks and the plant genetic resources community to the GBIF informatics infrastructure including the new toolkits for data exchange. This paper provides one of the first examples and guidelines for how to create extensions to the Darwin Core standard.

  1. The University of Missouri Research Reactor HEU to LEU conversion project status

    Energy Technology Data Exchange (ETDEWEB)

    McKibben, James C; Kutikkad, Kiratadas; Foyto, Leslie P; Peters, Nickie J; Solbrekken, Gary L; Kennedy, John [University of Missouri Research Reactor, Missouri (United States); Stillman, John A; Feldman, Earl E; Tzanos, Constantine P; Stevens, John G [Argonne National Laboratory, Argonne, Illinois (United States)

    2012-03-15

    The University of Missouri Research Reactor (MURR) is one of five U.S. high performance research and test reactors that are actively collaborating with the U.S. Department of Energy (DOE) to find a suitable low-enriched uranium (LEU) fuel replacement for the currently required highly-enriched uranium (HEU) fuel. A conversion feasibility study based on U-10Mo monolithic LEU fuel was completed in 2009. It was concluded that the proposed LEU fuel assembly design, in conjunction with an increase in power level from 10 to 12 MWth, will (1) maintain safety margins during operation, (2) allow operating fuel cycle lengths to be maintained for efficient and effective use of the facility, and (3) preserve an acceptable level and spectrum of key neutron fluxes to meet the scientific mission of the facility. The MURR and Argonne National Laboratory (ANL) team is continuing to work toward realization of the conversion. The 'Preliminary Safety Analysis Report Methodologies and Scenarios for LEU Conversion of MURR' was completed in June 2011. This report documents design parameter values critical to the Fuel Development (FD), Fuel Fabrication Capability (FFC) and Hydromechanical Fuel Test Facility (HMFTF) projects. The report also provides a preliminary evaluation of safety analysis techniques and data that will be needed to complete the fuel conversion Safety Analysis Report (SAR), especially those related to the U-10Mo monolithic LEU fuel. Specific studies are underway to validate the proposed path to an LEU fuel conversion. Coupled fluid-structure simulations and experiments are being conducted to understand the hydrodynamic plate deformation risk for 0.965 mm (38 mil) thick fuel plates. Methodologies that were recently developed to answer the U.S. Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) regarding the MURR 2006 relicensing submittal will be used in the LEU conversion effort. Transition LEU fuel elements that will have a minimal impact on

  2. [Study on standardization of cupping technique: elucidation on the establishment of the National Standard Standardized Manipulation of Acupuncture and Moxibustion, Part V, Cupping].

    Science.gov (United States)

    Gao, Shu-zhong; Liu, Bing

    2010-02-01

    From the aspects of basis, technique descriptions, core contents, problems and solutions, and standard thinking in standard setting process, this paper states experiences in the establishment of the national standard Standardized Manipulation of Acupuncture and Moxibustion, Part V, Cupping, focusing on methodologies used in cupping standard setting process, the method selection and operating instructions of cupping standardization, and the characteristics of standard TCM. In addition, this paper states the scope of application, and precautions for this cupping standardization. This paper also explaines tentative ideas on the research of standardized manipulation of acupuncture and moxibustion.

  3. Detection of cores in fingerprints with improved dimension reduction

    NARCIS (Netherlands)

    Bazen, A.M.; Veldhuis, Raymond N.J.

    In this paper, we present a statistical approach to core detection in fingerprint images that is based on the likelihood ratio, using models of variation of core templates and randomly chosen templates. Additionally, we propose an alternative dimension reduction method. Unlike standard linear

  4. Conversion program in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, E.B. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1997-08-01

    The conversion of the Swedish 50 MW R2 reactor from HEU to LEU fuel has been successfully accomplished over a 16 cycles long process. The conversion started in January 1991 with the introduction of 6 LEU assemblies in the 8*8 core. The first all LEU core was loaded in March 1993 and physics measurements were performed for the final licensing reports. A total of 142 LEU fuel assemblies have been irradiated up until September 1994 without any fuel incident. The operating licence for the R2 reactor was renewed in mid 1994 taking into account new fuel type. The Swedish Nuclear Inspectorate (SKI) pointed out one crucial problem with the LEU operation, that the back end of the LEU fuel cycle has not yet been solved. For the HEU fuel Sweden had the reprocessing alternative. The country is now relying heavily on the success of the USDOEs Off Site Fuels Policy to take back the spent fuel from the research reactors. They have in the meantime increased their intermediate storage facilities. There is, however, a limit both in time and space for storage of MTR-type of assemblies in water. The penalty of the lower thermal neutron flux in LEU cores has been reduced by improvements of the new irradiation rigs and by fine tuning the core calculations. The Studsvik code package, CASMO-SIMULATE, widely used for ICFM in LWRs has been modified to suit the compact MTR type of core.

  5. A study on improving the performance of a research reactor's equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2013-01-01

    Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

  6. Recommendation for measuring clinical outcome in distal radius fractures: a core set of domains for standardized reporting in clinical practice and research.

    Science.gov (United States)

    Goldhahn, Jörg; Beaton, Dorcas; Ladd, Amy; Macdermid, Joy; Hoang-Kim, Amy

    2014-02-01

    Lack of standardization of outcome measurement has hampered an evidence-based approach to clinical practice and research. We adopted a process of reviewing evidence on current use of measures and appropriate theoretical frameworks for health and disability to inform a consensus process that was focused on deriving the minimal set of core domains in distal radius fracture. We agreed on the following seven core recommendations: (1) pain and function were regarded as the primary domains, (2) very brief measures were needed for routine administration in clinical practice, (3) these brief measures could be augmented by additional measures that provide more detail or address additional domains for clinical research, (4) measurement of pain should include measures of both intensity and frequency as core attributes, (5) a numeric pain scale, e.g. visual analogue scale or visual numeric scale or the pain subscale of the patient-reported wrist evaluation (PRWE) questionnaires were identified as reliable, valid and feasible measures to measure these concepts, (6) for function, either the Quick Disability of the arm, shoulder and hand questionnaire or PRWE-function subscale was identified as reliable, valid and feasible measures, and (7) a measure of participation and treatment complications should be considered core outcomes for both clinical practice and research. We used a sound methodological approach to form a comprehensive foundation of content for outcomes in the area of distal radius fractures. We recommend the use of symptom and function as separate domains in the ICF core set in clinical research or practice for patients with wrist fracture. Further research is needed to provide more definitive measurement properties of measures across all domains.

  7. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    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 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. 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

  8. Irradiation tests of U3Si2-Al fuels up to very high fission densities

    International Nuclear Information System (INIS)

    Nuding, M.; Boening, K.

    2001-01-01

    The new research reactor of the Munich Technical University (TUM), the FRM-II, will have U 3 Si 2 -Al as the fuel. This fuel is considered qualified and optimally usable in the light of findings obtained in the RERTR program (Reduced Enrichment for Research and Test Reactors). The RERTR program was conducted to develop new fuel for the use of low enriched uranium (LEU) in research reactors. As the unique properties of the FRM-II in research and application are based also on achieving a very compact reactor core with highly enriched uranium (HEU), additional irradiation tests were performed on the basis of the RERTR program. They were run in close cooperation with the French Commissariat a l'Energie Atomique (CEA) in its SILOE and OSIRIS facilities, among others. After extensive evaluation, also of other studies, these tests confirm the RERTR findings about fuel swelling behavior and, consequently, the suitability of U 3 Si 2 -Al (HEU) for use in the compact core of the FRM-II. (orig.) [de

  9. The development of depletion program coupled with Monte Carlo computer code

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

    2015-01-01

    The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

  10. Core stability training on lower limb balance strength.

    Science.gov (United States)

    Dello Iacono, Antonio; Padulo, Johnny; Ayalon, Moshe

    2016-01-01

    This study aimed to assess the effects of core stability training on lower limbs' muscular asymmetries and imbalances in team sport. Twenty footballers were divided into two groups, either core stability or control group. Before each daily practice, core stability group (n = 10) performed a core stability training programme, while control group (n = 10) did a standard warm-up. The effects of the core stability training programme were assessed by performing isokinetic tests and single-leg countermovement jumps. Significant improvement was found for knee extensors peak torque at 3.14 rad · s(-1) (14%; P core stability group. The jump tests showed a significant reduction in the strength asymmetries in core stability group (-71.4%; P = 0.02) while a concurrent increase was seen in the control group (33.3%; P core exercises for optimal lower limbs strength balance development in young soccer players.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-01

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

  12. Improvement of core monitoring code cecor by the virtual segmentation of the self powered neutron detector loaded at Korean Standard Nuclear Plant

    International Nuclear Information System (INIS)

    Choi, T.; Jung, Y.S.

    2006-01-01

    Full text: Full text: Korean Standard Nuclear Plant uses Self Powered Neutron Detectors (SPNDs) to measure the neutron flux in the reactor core. The SPND's height is 40 cm and is located axially at the five different positions and 45 radial places. The design code simulated a reactor core is calculated by segmentation of the core. The segmentation is called as 'node', of which size is normally 20 cm. The axial height of the detector is larger than that of the node, and the larger detector's height maybe product some error on the axially complex shape. The analysis with the detector's signals showed some errors at the non-cosine axial flux shape. In order to reduce the errors for the shape, we tried to divide the detector by introducing the virtual boundary in the detector. Then, each axially 5 detectors had two virtual segmentations respectively and the detector's signal was divided by the inputs. So the more virtual detector's signals were gotten, the more accurate axial shape was produced. The result with virtual segmentations in a detector gave less deviation than the case without virtual segmentation (the current model). After the middle of cycle at the initial core specially, the axial neutron flux shape is changed to the saddle type one. The current model gave some error in Root Mean Square (RMS) between the measured value and the calculated one. The virtual segmentation model gave the better agreement at that time

  13. Some Main Results of Commissioning of the Dalat Research Reactor with Low Enriched Fuel

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2014-01-01

    After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements. (author)

  14. Setting standards to determine core clerkship grades in pediatrics.

    Science.gov (United States)

    Dudas, Robert A; Barone, Michael A

    2014-01-01

    One of the greatest challenges for clerkship directors is assigning a final grade and determining the precise point at which a student either passes or fails a clinical clerkship. The process of incorporating both subjective and objective assessment data to provide a final summative grade can be challenging. We describe our experience conducting a standard-setting exercise to set defensible cut points in a 4-tiered grading system in our pediatric clerkship. Using the Hofstee standard-setting approach, 8 faculty members participated in an exercise to establish grade cut points. These faculty members were subsequently surveyed to assess their attitudes toward the standard-setting process as well as their reactions to these newly proposed standards. We applied the new cut points to a historic cohort of 116 Johns Hopkins University School of Medicine students from the academic year 2012-2013 to assess the potential impact on grade distributions. The resultant grading schema would lead to a significant increase in the number of students receiving a failing grade and a decrease in the number of students receiving a grade of honors in a historical cohort. Faculty reported that the Hofstee method was easy to understand and fair. All faculty members thought that grade inflation presently exists within the pediatric clerkship. This study demonstrates that practical standards using the Hofstee method can be set for medical students in a pediatric clerkship in which multiple performance measures are used. Copyright © 2014 Academic Pediatric Association. Published by Elsevier Inc. All rights reserved.

  15. Role of ligand-ligand vs. core-core interactions in gold nanoclusters.

    Science.gov (United States)

    Milowska, Karolina Z; Stolarczyk, Jacek K

    2016-05-14

    The controlled assembly of ligand-coated gold nanoclusters (NCs) into larger structures paves the way for new applications ranging from electronics to nanomedicine. Here, we demonstrate through rigorous density functional theory (DFT) calculations employing novel functionals accounting for van der Waals forces that the ligand-ligand interactions determine whether stable assemblies can be formed. The study of NCs with different core sizes, symmetry forms, ligand lengths, mutual crystal orientations, and in the presence of a solvent suggests that core-to-core van der Waals interactions play a lesser role in the assembly. The dominant interactions originate from combination of steric effects, augmented by ligand bundling on NC facets, and related to them changes in electronic properties induced by neighbouring NCs. We also show that, in contrast to standard colloidal theory approach, DFT correctly reproduces the surprising experimental trends in the strength of the inter-particle interaction observed when varying the length of the ligands. The results underpin the importance of understanding NC interactions in designing gold NCs for a specific function.

  16. Evidence of fire resistance of hollow-core slabs

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl; Sørensen, Lars Schiøtt; Giuliani, Luisa

    is therefore going on in the Netherlands about the fire resistance of hollow-core slabs. In 2014 the producers of hollow-core slabs have published a report of a project called Holcofire containing a collection of 162 fire tests on hollow-core slabs giving for the first time an overview of the fire tests made....... The present paper analyses the evidence now available for assessment of the fire resistance of extruded hollow-core slabs. The 162 fire tests from the Holcofire report are compared against the requirements for testing from the product standard for hollow-core slabs EN1168 and knowledge about the possible......Hollow-core slabs have during the past 50 years comprised a variety of different structures with different cross-sections and reinforcement. At present the extruded hollow-core slabs without cross-reinforcement in the bottom flange and usually round or oval longitudinal channels (holes...

  17. Unallocated Off-Specification Highly Enriched Uranium: Recommendations for Disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bridges, D. N.; Boeke, S. G.; Tousley, D. R.; Bickford, W.; Goergen, C.; Williams, W.; Hassler, M.; Nelson, T.; Keck, R.; Arbital, J.

    2002-02-27

    The U.S. Department of Energy (DOE) has made significant progress with regard to disposition planning for 174 metric tons (MTU) of surplus Highly Enriched Uranium (HEU). Approximately 55 MTU of this 174 MTU are ''offspec'' HEU. (''Off-spec'' signifies that the isotopic or chemical content of the material does not meet the American Society for Testing and Materials standards for commercial nuclear reactor fuel.) Approximately 33 of the 55 MTU have been allocated to off-spec commercial reactor fuel per an Interagency Agreement between DOE and the Tennessee Valley Authority (1). To determine disposition plans for the remaining {approx}22 MTU, the DOE National Nuclear Security Administration (NNSA) Office of Fissile Materials Disposition (OFMD) and the DOE Office of Environmental Management (EM) co-sponsored this technical study. This paper represents a synopsis of the formal technical report (NNSA/NN-0014). The {approx} 22 MTU of off-spec HEU inventory in this study were divided into two main groupings: one grouping with plutonium (Pu) contamination and one grouping without plutonium. This study identified and evaluated 26 potential paths for the disposition of this HEU using proven decision analysis tools. This selection process resulted in recommended and alternative disposition paths for each group of HEU. The evaluation and selection of these paths considered criteria such as technical maturity, programmatic issues, cost, schedule, and environment, safety and health compliance. The primary recommendations from the analysis are comprised of 7 different disposition paths. The study recommendations will serve as a technical basis for subsequent programmatic decisions as disposition of this HEU moves into the implementation phase.

  18. Social Innovations vs international Trade? Core labour standards and exports

    OpenAIRE

    Siroën, Jean-Marc

    2011-01-01

    Labour standards defined by the ILO in 1998 are universal but applied very differently in countries. They are much better respected in high income countries. However, the causality between labour standards and growth remains a controversial issue. The strategies of export-led growth might encourage developing countries to contain the rising process of standards, first to increase their unskilled labour endowments for strengthening their comparative advantage relative to complying countries, a...

  19. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    Energy Technology Data Exchange (ETDEWEB)

    Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

  20. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    International Nuclear Information System (INIS)

    Wiencek, T.C.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

  1. Assessing the Genetics Content in the Next Generation Science Standards.

    Science.gov (United States)

    Lontok, Katherine S; Zhang, Hubert; Dougherty, Michael J

    2015-01-01

    Science standards have a long history in the United States and currently form the backbone of efforts to improve primary and secondary education in science, technology, engineering, and math (STEM). Although there has been much political controversy over the influence of standards on teacher autonomy and student performance, little light has been shed on how well standards cover science content. We assessed the coverage of genetics content in the Next Generation Science Standards (NGSS) using a consensus list of American Society of Human Genetics (ASHG) core concepts. We also compared the NGSS against state science standards. Our goals were to assess the potential of the new standards to support genetic literacy and to determine if they improve the coverage of genetics concepts relative to state standards. We found that expert reviewers cannot identify ASHG core concepts within the new standards with high reliability, suggesting that the scope of content addressed by the standards may be inconsistently interpreted. Given results that indicate that the disciplinary core ideas (DCIs) included in the NGSS documents produced by Achieve, Inc. clarify the content covered by the standards statements themselves, we recommend that the NGSS standards statements always be viewed alongside their supporting disciplinary core ideas. In addition, gaps exist in the coverage of essential genetics concepts, most worryingly concepts dealing with patterns of inheritance, both Mendelian and complex. Finally, state standards vary widely in their coverage of genetics concepts when compared with the NGSS. On average, however, the NGSS support genetic literacy better than extant state standards.

  2. Assessing the Genetics Content in the Next Generation Science Standards.

    Directory of Open Access Journals (Sweden)

    Katherine S Lontok

    Full Text Available Science standards have a long history in the United States and currently form the backbone of efforts to improve primary and secondary education in science, technology, engineering, and math (STEM. Although there has been much political controversy over the influence of standards on teacher autonomy and student performance, little light has been shed on how well standards cover science content. We assessed the coverage of genetics content in the Next Generation Science Standards (NGSS using a consensus list of American Society of Human Genetics (ASHG core concepts. We also compared the NGSS against state science standards. Our goals were to assess the potential of the new standards to support genetic literacy and to determine if they improve the coverage of genetics concepts relative to state standards. We found that expert reviewers cannot identify ASHG core concepts within the new standards with high reliability, suggesting that the scope of content addressed by the standards may be inconsistently interpreted. Given results that indicate that the disciplinary core ideas (DCIs included in the NGSS documents produced by Achieve, Inc. clarify the content covered by the standards statements themselves, we recommend that the NGSS standards statements always be viewed alongside their supporting disciplinary core ideas. In addition, gaps exist in the coverage of essential genetics concepts, most worryingly concepts dealing with patterns of inheritance, both Mendelian and complex. Finally, state standards vary widely in their coverage of genetics concepts when compared with the NGSS. On average, however, the NGSS support genetic literacy better than extant state standards.

  3. Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

    OpenAIRE

    Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

    2010-01-01

    Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

  4. Diagnostic accuracy of 22/25-gauge core needle in endoscopic ultrasound-guided sampling: systematic review and meta-analysis.

    Science.gov (United States)

    Oh, Hyoung-Chul; Kang, Hyun; Lee, Jae Young; Choi, Geun Joo; Choi, Jung Sik

    2016-11-01

    To compare the diagnostic accuracy of endoscopic ultrasound-guided core needle aspiration with that of standard fine-needle aspiration by systematic review and meta-analysis. Studies using 22/25-gauge core needles, irrespective of comparison with standard fine needles, were comprehensively reviewed. Pooled sensitivity, specificity, diagnostic odds ratio (DOR), and summary receiver operating characteristic curves for the diagnosis of malignancy were used to estimate the overall diagnostic efficiency. The pooled sensitivity, specificity, and DOR of the core needle for the diagnosis of malignancy were 0.88 (95% confidence interval [CI], 0.84 to 0.90), 0.99 (95% CI, 0.96 to 1), and 167.37 (95% CI, 65.77 to 425.91), respectively. The pooled sensitivity, specificity, and DOR of the standard needle were 0.84 (95% CI, 0.79 to 0.88), 1 (95% CI, 0.97 to 1), and 130.14 (95% CI, 34.00 to 495.35), respectively. The area under the curve of core and standard needle in the diagnosis of malignancy was 0.974 and 0.955, respectively. The core and standard needle were comparable in terms of pancreatic malignancy diagnosis. There was no significant difference in procurement of optimal histologic cores between core and standard needles (risk ratio [RR], 0.545; 95% CI, 0.187 to 1.589). The number of needle passes for diagnosis was significantly lower with the core needle (standardized mean difference, -0.72; 95% CI, -1.02 to -0.41). There were no significant differences in overall complications (RR, 1.26; 95% CI, 0.34 to 4.62) and technical failure (RR, 5.07; 95% CI, 0.68 to 37.64). Core and standard needles were comparable in terms of diagnostic accuracy, technical performance, and safety profile.

  5. Evaluation of BEACON-COLSS Core Monitoring System Benefits

    International Nuclear Information System (INIS)

    Kim, Joon Sung; Park, Young Ho; Morita, Toshio; Book, Michael A.

    2005-01-01

    In Korean Standard Nuclear Power Plant COLSS (Core Operating Limit Supervisory System) is used to monitor the DNBR Power Operating Limit (DNBRPOL) and Linear Heat Rate POL (KWPFPOL). Westinghouse and KNFC have developed an upgraded core monitoring system by combining the BEACON TM core monitoring system 1 (Best Estimate Analyzer for Core Operation . Nuclear) and COLSS into an integrated product that is called BEACON-COLSS. BEACON-COLSS generates the 3-D power distribution corrected by the in-core detectors measurements. The 3-D core power distribution methodology in BEACON-COLSS is significantly better than the synthesis methodology in COLSS. BEACONCOLSS uses the CETOP-D 2 thermal hydraulic code instead of CETOP-1. CETOP-D is a multi-channel thermal hydraulics code that will provide more accurate DNBR calculations than the DNBR calculators currently used in COLSS

  6. Management of high enriched uranium for peaceful purposes: Status and trends

    International Nuclear Information System (INIS)

    2005-06-01

    Arms control agreements between some Nuclear Weapon States have led to the dismantling of many of the nuclear weapons in their military stockpiles, which in turn have produced stockpiles of excess weapons-grade high enriched uranium (HEU) from the dismantled weapons. Considering the proliferation potential of HEU, the management, control and disposition of this fissile material has become a primary focus of nuclear non-proliferation efforts worldwide. To lessen the proliferation threat of excess HEU stockpiles, the USA agreed to purchase several tonnes of excess Russian HEU down-blended to low enriched uranium (LEU). Proliferation concerns about HEU have also resulted in a global effort to convert research reactors from HEU to LEU fuel and to minimize civilian use of HEU. This publication addresses HEU management declared excesses, non-proliferation programmes and options for the use of HEU stockpiles, including disposition programmes. Also addressed are the influence of LEU derived from surplus HEU on the global market for uranium, technical issues associated with utilization and the disposition of HEU

  7. Progress in standardization for ITER Remote Handling control system

    International Nuclear Information System (INIS)

    Hamilton, David Thomas; Tesini, Alessandro; Ranz, Roberto; Kozaka, Hiroshi

    2014-01-01

    Graphical abstract: - Highlights: • Standard parts specified for ITER Remote Handling (RH) control system. • Standard approach for VR modeling of structural deformations in real-time. • RH Core System produced as standard platform for RH controller applications. • Synthetic Viewing investigated and demonstrated. • Structured language defined for RH operation procedures and motion sequences. - Abstract: An integrated control system architecture has been defined for the ITER Remote Handling (RH) equipment systems, and work has been continuing to develop and validate standards for this architecture. Evaluations of standard parts and a standard control room work-cell have contributed to an update of the RH Control System Design Handbook, while R and D activities have been carried out to validate concepts for standard solutions to ITER RH problems: the use of a standard master arm with different slave arms, the achievement of high accuracy tracking of RH operations within virtual reality, and condition monitoring of RH equipment systems. The standardization efforts have been consolidated through the development of a freely distributable software platform to support the adoption of the ITER RH standards. The RH Core System installs on top of the CODAC Core System and provides the basic platform for the development of ITER RH equipment controller applications. The standardization work has continued in the areas of RH viewing, network communication protocols, and a structured language for programming ITER RH operations. Prototyping has been done on high-level control system applications, and R and D has been carried out in the area of synthetic viewing for ITER RH. These developments will be reflected in a new version of the RH Core System to be produced during 2013

  8. Progress in standardization for ITER Remote Handling control system

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, David Thomas, E-mail: david.hamilton@iter.org [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Tesini, Alessandro [ITER Organization, Route de Vinon, 13115 St. Paul-lez-Durance (France); Ranz, Roberto [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Kozaka, Hiroshi [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan)

    2014-10-15

    Graphical abstract: - Highlights: • Standard parts specified for ITER Remote Handling (RH) control system. • Standard approach for VR modeling of structural deformations in real-time. • RH Core System produced as standard platform for RH controller applications. • Synthetic Viewing investigated and demonstrated. • Structured language defined for RH operation procedures and motion sequences. - Abstract: An integrated control system architecture has been defined for the ITER Remote Handling (RH) equipment systems, and work has been continuing to develop and validate standards for this architecture. Evaluations of standard parts and a standard control room work-cell have contributed to an update of the RH Control System Design Handbook, while R and D activities have been carried out to validate concepts for standard solutions to ITER RH problems: the use of a standard master arm with different slave arms, the achievement of high accuracy tracking of RH operations within virtual reality, and condition monitoring of RH equipment systems. The standardization efforts have been consolidated through the development of a freely distributable software platform to support the adoption of the ITER RH standards. The RH Core System installs on top of the CODAC Core System and provides the basic platform for the development of ITER RH equipment controller applications. The standardization work has continued in the areas of RH viewing, network communication protocols, and a structured language for programming ITER RH operations. Prototyping has been done on high-level control system applications, and R and D has been carried out in the area of synthetic viewing for ITER RH. These developments will be reflected in a new version of the RH Core System to be produced during 2013.

  9. Core strength training for patients with chronic low back pain

    OpenAIRE

    Chang, Wen-Dien; Lin, Hung-Yu; Lai, Ping-Tung

    2015-01-01

    [Purpose] Through core strength training, patients with chronic low back pain can strengthen their deep trunk muscles. However, independent training remains challenging, despite the existence of numerous core strength training strategies. Currently, no standardized system has been established analyzing and comparing the results of core strength training and typical resistance training. Therefore, we conducted a systematic review of the results of previous studies to explore the effectiveness ...

  10. Abstracts and papers of the 2002 International RERTR Meeting

    International Nuclear Information System (INIS)

    2002-11-01

    This 24th RERTR meeting has been organized in cooperation with the IAEA. It was attended by almost 200 participants coming from almost 30 different countries. Among them we have representatives from governmental and international organizations, private companies, reactor operators, fuel developers, regulators. The papers presented during following sessions: national and international RERTR programs, fuel development, manufacturing and qualification; fuel performance; HEU and LEU fuel cycle; LEU target development; licensing, safety and core assessment; reactor core conversion; acceptance program and spent fuel transportation; spent fuel management

  11. Tidal disruption of fuzzy dark matter subhalo cores

    Science.gov (United States)

    Du, Xiaolong; Schwabe, Bodo; Niemeyer, Jens C.; Bürger, David

    2018-03-01

    We study tidal stripping of fuzzy dark matter (FDM) subhalo cores using simulations of the Schrödinger-Poisson equations and analyze the dynamics of tidal disruption, highlighting the differences with standard cold dark matter. Mass loss outside of the tidal radius forces the core to relax into a less compact configuration, lowering the tidal radius. As the characteristic radius of a solitonic core scales inversely with its mass, tidal stripping results in a runaway effect and rapid tidal disruption of the core once its central density drops below 4.5 times the average density of the host within the orbital radius. Additionally, we find that the core is deformed into a tidally locked ellipsoid with increasing eccentricities until it is completely disrupted. Using the core mass loss rate, we compute the minimum mass of cores that can survive several orbits for different FDM particle masses and compare it with observed masses of satellite galaxies in the Milky Way.

  12. Fire resistance of extruded hollow-core slabs

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl; Giuliani, Luisa; Sørensen, Lars Schiøtt

    2016-01-01

    Prefabricated extruded hollow-core slabs are preferred building components for floor structures in several countries. It is therefore important to be able to document the fire resistance of these slabs proving fulfilment of standard fire resistance requirements of 60- and 120 minutes found in most...... a standard fire test and from a thorough examination of the comprehensive test documentation available on fire exposed hollow-core slabs. Mechanisms for loss of load-bearing capacity are clarified, and evidence of the fire resistance is found. For the first time the mechanisms responsible for loss of load......-bearing capacity are identified and test results and calculation approach are for the first time Applied in accordance with each other for assessment of fire resistance of the structure....

  13. caCORE: a common infrastructure for cancer informatics.

    Science.gov (United States)

    Covitz, Peter A; Hartel, Frank; Schaefer, Carl; De Coronado, Sherri; Fragoso, Gilberto; Sahni, Himanso; Gustafson, Scott; Buetow, Kenneth H

    2003-12-12

    Sites with substantive bioinformatics operations are challenged to build data processing and delivery infrastructure that provides reliable access and enables data integration. Locally generated data must be processed and stored such that relationships to external data sources can be presented. Consistency and comparability across data sets requires annotation with controlled vocabularies and, further, metadata standards for data representation. Programmatic access to the processed data should be supported to ensure the maximum possible value is extracted. Confronted with these challenges at the National Cancer Institute Center for Bioinformatics, we decided to develop a robust infrastructure for data management and integration that supports advanced biomedical applications. We have developed an interconnected set of software and services called caCORE. Enterprise Vocabulary Services (EVS) provide controlled vocabulary, dictionary and thesaurus services. The Cancer Data Standards Repository (caDSR) provides a metadata registry for common data elements. Cancer Bioinformatics Infrastructure Objects (caBIO) implements an object-oriented model of the biomedical domain and provides Java, Simple Object Access Protocol and HTTP-XML application programming interfaces. caCORE has been used to develop scientific applications that bring together data from distinct genomic and clinical science sources. caCORE downloads and web interfaces can be accessed from links on the caCORE web site (http://ncicb.nci.nih.gov/core). caBIO software is distributed under an open source license that permits unrestricted academic and commercial use. Vocabulary and metadata content in the EVS and caDSR, respectively, is similarly unrestricted, and is available through web applications and FTP downloads. http://ncicb.nci.nih.gov/core/publications contains links to the caBIO 1.0 class diagram and the caCORE 1.0 Technical Guide, which provide detailed information on the present caCORE architecture

  14. Supply of enriched uranium for research reactors

    International Nuclear Information System (INIS)

    Mueller, Hans; Laucht, Juergen

    1996-01-01

    Since the RERTR meeting in 1990 at Newport/USA, NUKEM recommended that the research reactor community agree upon a worldwide unified technical specification for low enriched uranium (LEU) and high enriched uranium (HEU) since there existed numerous specifications both from suppliers/fabricators and research reactors. The target recommended by NUKEM is to arrive at a worldwide unified standard specification in order to facilitate supplies of LEU and HEU to fabricators for fabrication of research reactor fuel elements. In our paper presented at the RERTR meeting at Paris in September 1995, we pointed out that LEU and HEU supplied by the U.S. Department of Energy (DOE) in the past was never 'virgin' material, i.e., it was mixed with reprocessed uranium. Our recommendation was to include this fact in the proposed unified specification. Since the RERTR meeting in 1995 progress on a unified standard specification has been made and we would like to provide more specific information about that in this paper. Furthermore, we will deal with the question whether there is a secure supply of LEU for converted research reactors. We list current and potential suppliers of LEU, noting however, that the DOE has for a number of years been unable to supply any LEU due to production problems. The future availability of LEU of U.S. origin is, of course, essential for those research reactor operators which have converted their reactors from HEU to LEU and which are intending to return spent fuel of U.S. origin to the U.S.A. (author)

  15. Core Certification of Data Repositories: Trustworthiness and Long-Term Stewardship

    Science.gov (United States)

    de Sherbinin, A. M.; Mokrane, M.; Hugo, W.; Sorvari, S.; Harrison, S.

    2017-12-01

    Scientific integrity and norms dictate that data created and used by scientists should be managed, curated, and archived in trustworthy data repositories thus ensuring that science is verifiable and reproducible while preserving the initial investment in collecting data. Research stakeholders including researchers, science funders, librarians, and publishers must also be able to establish the trustworthiness of data repositories they use to confirm that the data they submit and use remain useful and meaningful in the long term. Data repositories are increasingly recognized as a key element of the global research infrastructure and the importance of establishing their trustworthiness is recognised as a prerequisite for efficient scientific research and data sharing. The Core Trustworthy Data Repository Requirements are a set of universal requirements for certification of data repositories at the core level (see: https://goo.gl/PYsygW). They were developed by the ICSU World Data System (WDS: www.icsu-wds.org) and the Data Seal of Approval (DSA: www.datasealofapproval.org)—the two authoritative organizations responsible for the development and implementation of this standard to be further developed under the CoreTrustSeal branding . CoreTrustSeal certification of data repositories involves a minimally intensive process whereby repositories supply evidence that they are sustainable and trustworthy. Repositories conduct a self-assessment which is then reviewed by community peers. Based on this review CoreTrustSeal certification is granted by the CoreTrustSeal Standards and Certification Board. Certification helps data communities—producers, repositories, and consumers—to improve the quality and transparency of their processes, and to increase awareness of and compliance with established standards. This presentation will introduce the CoreTrustSeal certification requirements for repositories and offer an opportunity to discuss ways to improve the contribution of

  16. Current status of core needle biopsy of the thyroid

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jung Hwan [Dept. of Radiology and Research Institute of Radiology, Asan Medical Center, University of Ulsan College of Medicine, Seoul (Korea, Republic of)

    2017-04-15

    Thyroid nodules are a common clinical problem. Fine-needle aspiration (FNA) and large-needle biopsy have been used to diagnose thyroid nodules. Before the 1980s, large-needle biopsy was the standard procedure for the thyroid, but FNA became the standard diagnostic tool in the 1980s because it is a safe procedure that leads to accurate diagnoses. With advances in core needle biopsy (CNB) devices (i.e., spring-activated core needles) and development of high-resolution ultrasound, it has become possible to make accurate diagnoses while minimizing complications. Although 18- to 21-gauge core needles can be used to biopsy thyroid nodules, 18-gauge needles are most commonly used in Korea. The relationships among the size of the needle, the number of core specimens, and diagnostic accuracy have not yet been conclusively established, but the general tendency is that thinner needles cause less damage to the normal thyroid, but allow a smaller amount of thyroid tissue to be biopsied to be obtained. These relationships may be validated in the future.

  17. Chilean fuel elements fabrication progress report

    International Nuclear Information System (INIS)

    Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

    1993-01-01

    Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

  18. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    Energy Technology Data Exchange (ETDEWEB)

    Nagaoka, Yoshiharu; Oyamada, Rokuro [Japan Atomic Energy Research Institute, Oarai-machi Ibaraki-ken (Japan); Matos, J E; Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

  19. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    International Nuclear Information System (INIS)

    Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

    1985-01-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

  20. The Expanded FindCore Method for Identification of a Core Atom Set for Assessment of Protein Structure Prediction

    Science.gov (United States)

    Snyder, David A.; Grullon, Jennifer; Huang, Yuanpeng J.; Tejero, Roberto; Montelione, Gaetano T.

    2014-01-01

    Maximizing the scientific impact of NMR-based structure determination requires robust and statistically sound methods for assessing the precision of NMR-derived structures. In particular, a method to define a core atom set for calculating superimpositions and validating structure predictions is critical to the use of NMR-derived structures as targets in the CASP competition. FindCore (D.A. Snyder and G.T. Montelione PROTEINS 2005;59:673–686) is a superimposition independent method for identifying a core atom set, and partitioning that set into domains. However, as FindCore optimizes superimposition by sensitively excluding not-well-defined atoms, the FindCore core may not comprise all atoms suitable for use in certain applications of NMR structures, including the CASP assessment process. Adapting the FindCore approach to assess predicted models against experimental NMR structures in CASP10 required modification of the FindCore method. This paper describes conventions and a standard protocol to calculate an “Expanded FindCore” atom set suitable for validation and application in biological and biophysical contexts. A key application of the Expanded FindCore method is to identify a core set of atoms in the experimental NMR structure for which it makes sense to validate predicted protein structure models. We demonstrate the application of this Expanded FindCore method in characterizing well-defined regions of 18 NMR-derived CASP10 target structures. The Expanded FindCore protocol defines “expanded core atom sets” that match an expert’s intuition of which parts of the structure are sufficiently well-defined to use in assessing CASP model predictions. We also illustrate the impact of this analysis on the CASP GDT assessment scores. PMID:24327305

  1. Air core detectors for Cerenkov-free scintillation dosimetry of brachytherapy β-sources.

    Science.gov (United States)

    Eichmann, Marion; Thomann, Benedikt

    2017-09-01

    Plastic scintillation detectors are used for dosimetry in small radiation fields with high dose gradients, e.g., provided by β-emitting sources like 106 Ru/ 106 Rh eye plaques. A drawback is a background signal caused by Cerenkov radiation generated by electrons passing the optical fibers (light guides) of this dosimetry system. Common approaches to correct for the Cerenkov signal are influenced by uncertainties resulting from detector positioning and calibration procedures. A different approach to avoid any correction procedure is to suppress the Cerenkov signal by replacing the solid core optical fiber with an air core light guide, previously shown for external beam therapy. In this study, the air core concept is modified and applied to the requirements of dosimetry in brachytherapy, proving its usability for measuring water energy doses in small radiation fields. Three air core detectors with different air core lengths are constructed and their performance in dosimetry for brachytherapy β-sources is compared with a standard two-fiber system, which uses a second fiber for Cerenkov correction. The detector systems are calibrated with a 90 Sr/ 90 Y secondary standard and tested for their angular dependence as well as their performance in depth dose measurements of 106 Ru/ 106 Rh sources. The signal loss relative to the standard detector increases with increasing air core length to a maximum value of 58.3%. At the same time, however, the percentage amount of Cerenkov light in the total signal is reduced from at least 12.1% to a value below 1.1%. There is a linear correlation between induced dose and measured signal current. The air core detectors determine the dose rates for 106 Ru/ 106 Rh sources without any form of correction for the Cerenkov signal. The air core detectors show advantages over the standard two-fiber system especially when measuring in radiation fields with high dose gradients. They can be used as simple one-fiber systems and allow for an almost

  2. Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

    International Nuclear Information System (INIS)

    Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

    2003-01-01

    The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

  3. The core content of clinical ultrasonography fellowship training.

    Science.gov (United States)

    Lewiss, Resa E; Tayal, Vivek S; Hoffmann, Beatrice; Kendall, John; Liteplo, Andrew S; Moak, James H; Panebianco, Nova; Noble, Vicki E

    2014-04-01

    The purpose of developing a core content for subspecialty training in clinical ultrasonography (US) is to standardize the education and qualifications required to provide oversight of US training, clinical use, and administration to improve patient care. This core content would be mastered by a fellow as a separate and unique postgraduate training, beyond that obtained during an emergency medicine (EM) residency or during medical school. The core content defines the training parameters, resources, and knowledge of clinical US necessary to direct clinical US divisions within medical specialties. Additionally, it is intended to inform fellowship directors and candidates for certification of the full range of content that might appear in future examinations. This article describes the development of the core content and presents the core content in its entirety. © 2014 by the Society for Academic Emergency Medicine.

  4. Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems

    International Nuclear Information System (INIS)

    1994-12-01

    The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors

  5. Core Standards of the EUBIROD Project. Defining a European Diabetes Data Dictionary for Clinical Audit and Healthcare Delivery.

    Science.gov (United States)

    Cunningham, S G; Carinci, F; Brillante, M; Leese, G P; McAlpine, R R; Azzopardi, J; Beck, P; Bratina, N; Bocquet, V; Doggen, K; Jarosz-Chobot, P K; Jecht, M; Lindblad, U; Moulton, T; Metelko, Ž; Nagy, A; Olympios, G; Pruna, S; Skeie, S; Storms, F; Di Iorio, C T; Massi Benedetti, M

    2016-01-01

    A set of core diabetes indicators were identified in a clinical review of current evidence for the EUBIROD project. In order to allow accurate comparisons of diabetes indicators, a standardised currency for data storage and aggregation was required. We aimed to define a robust European data dictionary with appropriate clinical definitions that can be used to analyse diabetes outcomes and provide the foundation for data collection from existing electronic health records for diabetes. Existing clinical datasets used by 15 partner institutions across Europe were collated and common data items analysed for consistency in terms of recording, data definition and units of measurement. Where necessary, data mappings and algorithms were specified in order to allow partners to meet the standard definitions. A series of descriptive elements were created to document metadata for each data item, including recording, consistency, completeness and quality. While datasets varied in terms of consistency, it was possible to create a common standard that could be used by all. The minimum dataset defined 53 data items that were classified according to their feasibility and validity. Mappings and standardised definitions were used to create an electronic directory for diabetes care, providing the foundation for the EUBIROD data analysis repository, also used to implement the diabetes registry and model of care for Cyprus. The development of data dictionaries and standards can be used to improve the quality and comparability of health information. A data dictionary has been developed to be compatible with other existing data sources for diabetes, within and beyond Europe.

  6. Metaphors We Do Math By: A Comparative Case Study of Public and Catholic School Teachers’ Understanding of the Common Core State Standards in Mathematics

    OpenAIRE

    Branch, Jennifer Danielle

    2016-01-01

    The United States has undergone multiple mathematics reforms since the 1980s with each reform setting out to increase national test scores and improve mathematics education in the nation’s schools. The current reform, the Common Core State Standards for Mathematics (CCSSM), seeks to create mathematically proficient students through a more active and rigorous curriculum. The goal of this yearlong study was to examine the understanding that intermediate and middle school math teachers make of t...

  7. Characterization analysis database system (CADS). A system overview

    International Nuclear Information System (INIS)

    1997-12-01

    The CADS database is a standardized, quality-assured, and configuration-controlled data management system developed to assist in the task of characterizing the DOE surplus HEU material. Characterization of the surplus HEU inventory includes identifying the specific material; gathering existing data about the inventory; defining the processing steps that may be necessary to prepare the material for transfer to a blending site; and, ultimately, developing a range of the preliminary cost estimates for those processing steps. Characterization focuses on producing commercial reactor fuel as the final step in material disposition. Based on the project analysis results, the final determination will be made as to the viability of the disposition path for each particular item of HEU. The purpose of this document is to provide an informational overview of the CADS database, its evolution, and its current capabilities. This document describes the purpose of CADS, the system requirements it fulfills, the database structure, and the operational guidelines of the system

  8. Nuclear Data Center International Standard Towards TSO Initiative

    International Nuclear Information System (INIS)

    Raja Murzaferi Raja Moktar; Mohd Fauzi Haris; Siti Nurbahyah Hamdan

    2011-01-01

    Nuclear Data Center is the main facility for Nuclear Malaysia Agency IT infrastructure comprising of main critical servers, research and operational data storage, HPC-clusters system and vital network core equipment. In recent years, international body such as TIA-Telecommunication Industry Association and Up time Institute have came out with proper international data center standards in order to ensure data center operation on achieving maximum operational up time and minimal downtime. The standard are currently being rated as tier level ranging from Data Center tier I up to tier IV, differentiate by facility standard and up time/ downtime percentage ratio. This paper will discuss Nuclear Data Center adopting international standards in supporting Nuclear Malaysia TSO initiative thus ensuring the critical core component of agency IT services availability and further more International standard recognitions. (author)

  9. DER NEUE STANDARD IFRS 15 - IASB UND FASB VERABSCHIEDEN EINEN WEITGEHEND EINHEITLICHEN STANDARD ZUR UMSATZREALISIERUNG

    OpenAIRE

    Bodo Runzheimer

    2015-01-01

    The international accounting regulations regarding sales revenues have been changed. In May 2014, the International Accounting Standards Board (IASB) and the US regulator, the Financial Accounting Standards Board (FASB), published together new revenue recording regulations, which will be applied in IFRS (International Financial Reporting Standards) as well as in US-GAAP (United States Generally Accepted Accounting Principles). The core principle of the new IFRS 15 is that an entity will recog...

  10. Status of reduced enrichment program for research reactors in Japan

    International Nuclear Information System (INIS)

    Kaieda, Keisuke; Baba, Osamu; Nagaoka, Yoshiharu; Kanda, Keiji; Nakagome, Yoshihiro

    1999-01-01

    The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI) have been completed. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M commenced using LEU silicide fuel elements instead of LEU aluminide fuel elements in September, 1999. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and April 1994 the U.S. Government gave an approval to utilize HEU fuel in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until March 2004, then the full core conversion with LEU silicide will be done. The first shipment of spent fuels since 1974 was done in August, 1999. (author)

  11. Ky. Road-Tests Common Core

    Science.gov (United States)

    Ujifusa, Andrew

    2012-01-01

    Results from new state tests in Kentucky--the first in the nation explicitly tied to the Common Core State Standards--show that the share of students scoring "proficient" or better in reading and math dropped by roughly a third or more in both elementary and middle school the first year the tests were given. Kentucky in 2010 was the…

  12. An alternative LEU design for the FRM-II

    International Nuclear Information System (INIS)

    Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

    1996-01-01

    The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance. LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. The following issues raised by TUM were addressed in Ref. 1: qualification of HEU and LEU silicide fuels, gamma heating in the heavy water reflector, radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. The conclusions of these analyses are summarized below. This paper addresses three additional safety issues that were raised by TUM in Ref. 2: stability of the involute fuel plates, a hypothetical accident involving the configuration of the reflector, and a loss of primary coolant flow transient due to an interrupted power supply. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

  13. Integration of Biosafety into Core Facility Management

    Science.gov (United States)

    Fontes, Benjamin

    2013-01-01

    This presentation will discuss the implementation of biosafety policies for small, medium and large core laboratories with primary shared objectives of ensuring the control of biohazards to protect core facility operators and assure conformity with applicable state and federal policies, standards and guidelines. Of paramount importance is the educational process to inform core laboratories of biosafety principles and policies and to illustrate the technology and process pathways of the core laboratory for biosafety professionals. Elevating awareness of biohazards and the biosafety regulatory landscape among core facility operators is essential for the establishment of a framework for both project and material risk assessment. The goal of the biohazard risk assessment process is to identify the biohazard risk management parameters to conduct the procedure safely and in compliance with applicable regulations. An evaluation of the containment, protective equipment and work practices for the procedure for the level of risk identified is facilitated by the establishment of a core facility registration form for work with biohazards and other biological materials with potential risk. The final step in the biocontainment process is the assumption of Principal Investigator role with full responsibility for the structure of the site-specific biosafety program plan by core facility leadership. The presentation will provide example biohazard protocol reviews and accompanying containment measures for core laboratories at Yale University.

  14. Neutronic analysis of the JMTR with LEU fuel and burnable poison

    International Nuclear Information System (INIS)

    Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

    1984-01-01

    The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed. 2 refs., 10 figs., 5 tabs

  15. Creating Next Generation Teacher Preparation Programs to Support Implementation of the Next Generation Science Standards and Common Core State Standards in K-12 Schools: An Opportunity for the Earth and Space Sciences

    Science.gov (United States)

    Geary, E. E.; Egger, A. E.; Julin, S.; Ronca, R.; Vokos, S.; Ebert, E.; Clark-Blickenstaff, J.; Nollmeyer, G.

    2015-12-01

    A consortium of two and four year Washington State Colleges and Universities in partnership with Washington's Office of the Superintendent of Public Instruction (OSPI), the Teachers of Teachers of Science, and Teachers of Teachers of Mathematics, and other key stakeholders, is currently working to improve science and mathematics learning for all Washington State students by creating a new vision for STEM teacher preparation in Washington State aligned with the Next Generation Science Standards (NGSS) and the Common Core State Standards (CCSS) in Mathematics and Language Arts. Specific objectives include: (1) strengthening elementary and secondary STEM Teacher Preparation courses and curricula, (2) alignment of STEM teacher preparation programs across Washington State with the NGSS and CCSS, (3) development of action plans to support implementation of STEM Teacher Preparation program improvement at Higher Education Institutions (HEIs) across the state, (4) stronger collaborations between HEIs, K-12 schools, government agencies, Non-Governmental Organizations, and STEM businesses, involved in the preparation of preservice STEM teachers, (5) new teacher endorsements in Computer Science and Engineering, and (6) development of a proto-type model for rapid, adaptable, and continuous improvement of STEM teacher preparation programs. A 2015 NGSS gap analysis of teacher preparation programs across Washington State indicates relatively good alignment of courses and curricula with NGSS Disciplinary Core Ideas and Scientific practices, but minimal alignment with NGSS Engineering practices and Cross Cutting Concepts. Likewise, Computer Science and Sustainability ideas and practices are not well represented in current courses and curricula. During the coming year teams of STEM faculty, education faculty and administrators will work collaboratively to develop unique action plans for aligning and improving STEM teacher preparation courses and curricula at their institutions.

  16. Web-based Core Design System Development

    International Nuclear Information System (INIS)

    Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

    2011-01-01

    The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

  17. Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium in Mixed Oxide

    International Nuclear Information System (INIS)

    Mount, M.; O'Connell, W.; Cochran, C.; Rinard, P.; Dearborn, D.; Endres, E.

    2002-01-01

    As a follow-on to the Lawrence Livermore National Laboratory (LLNL) effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler for measurement of highly enriched uranium (HEU) oxide, a method has been developed to extend the use of the PAN shuffler to the measurement of HEU in mixed uranium-plutonium (U-Pu) oxide. This method uses the current LLNL HEU oxide calibration algorithms, appropriately corrected for the mixed U-Pu oxide assay time, and recently developed PuO 2 calibration algorithms to yield the mass of 235 U present via differences between the expected count rate for the PuO 2 and the measured count rate of the mixed U-Pu oxide. This paper describes the LLNL effort to use PAN shuffler measurements of units of certified reference material (CRM) 149 (uranium (93% Enriched) Oxide - U 3 O 8 Standard for Neutron Counting Measurements) and CRM 146 (uranium Isotopic Standard for Gamma Spectrometry Measurements) and a selected set of LLNL PuO 2 -bearing containers in consort with Monte Carlo simulations of the PAN shuffler response to each to (1) establish and validate a correction to the HEU calibration algorithm for the mixed U-Pu oxide assay time, (2) develop a PuO 2 calibration algorithm that includes the effect of PuO 2 density (2.4 g/cm 3 to 4.8 g/cm 3 ) and container size (8.57 cm to 9.88 cm inside diameter and 9.60 cm to 13.29 cm inside height) on the PAN shuffler response, and (3) develop and validate the method for establishing the mass of 235 U present in an unknown of mixed U-Pu oxide.

  18. Beacon-Colss core monitoring system application and benefits

    International Nuclear Information System (INIS)

    Boyd, W.A.; Yoon, T.Y.

    2005-01-01

    Westinghouse and KNFC are creating an upgraded core monitoring system by merging the BEACON system (best estimate analyzer for core operation-nuclear) and COLSS (core operating limit supervisory system) into an integrated product. Although both BEACON and COLSS are core monitoring systems that have been in operation at many plants for a number of years, they each have some features and capabilities that are not in the other. Therefore it has been decided to incorporate portions of COLSS into the beacon system to create an optional level to support core monitoring applications on selected combustion engineering (C-E) designed plants. This optional level in the beacon system will be called BEACON-COLSS and will allow the beacon system to monitor the LCO's and Tech Spec limits at CE plants that currently use COLSS. This paper will present the structure of the new core monitoring system and the benefits it achieves for current COLSS plants, i.e., CE plants in the US and KSNP (Korean standard nuclear power plant). (authors)

  19. Determining the dark matter mass with DeepCore

    Energy Technology Data Exchange (ETDEWEB)

    Das, Chitta R. [Centro de Física Teórica de Partículas, Instituto Superior Técnico (CFTP), Universidade Tćnica de Lisboa, Avenida Rovisco Pais 1, 1049-001 Lisboa (Portugal); Mena, Olga [Instituto de Física Corpuscular (IFIC), CSIC-Universitat de València, Apartado de Correos 22085, E-46071 Valencia (Spain); Palomares-Ruiz, Sergio, E-mail: sergio.palomares.ruiz@ist.utl.pt [Centro de Física Teórica de Partículas, Instituto Superior Técnico (CFTP), Universidade Tćnica de Lisboa, Avenida Rovisco Pais 1, 1049-001 Lisboa (Portugal); Instituto de Física Corpuscular (IFIC), CSIC-Universitat de València, Apartado de Correos 22085, E-46071 Valencia (Spain); Pascoli, Silvia [IPPP, Department of Physics, Durham University, Durham DH1 3LE (United Kingdom)

    2013-10-01

    Cosmological and astrophysical observations provide increasing evidence of the existence of dark matter in our Universe. Dark matter particles with a mass above a few GeV can be captured by the Sun, accumulate in the core, annihilate, and produce high energy neutrinos either directly or by subsequent decays of Standard Model particles. We investigate the prospects for indirect dark matter detection in the IceCube/DeepCore neutrino telescope and its capabilities to determine the dark matter mass.

  20. Does the Common Core Further Democracy? A Response to "The Common Core and Democratic Education: Examining Potential Costs and Benefits to Public and Private Autonomy"

    Science.gov (United States)

    Neem, Johann N.

    2018-01-01

    The Common Core does not advance democratic education. Far from it, the opening section of the language standards argues that the goal of public K-12 education is "college and career readiness." Only at the end of their introductory section do the Common Core's authors suggest that K-12 education has any goals beyond the economic:…

  1. Expert's statement on the research reactor Munich II (FRM-II); Gutachterliche Stellungnahme zum Forschungsreaktor Muenchen II (FRM-II)

    Energy Technology Data Exchange (ETDEWEB)

    Liebert, Wolfgang; Friess, Friederike; Gufler, Klaus; Arnold, Nikolaus [Univ. fuer Bodenkultur (BOKU), Wien (Austria). Inst. fuer Sicherheits- und Risikowissenschaften (ISR)

    2017-12-15

    The Expert's statement on the research reactor FRM-II covers the following issues: The situation in Germany with respect to HEU (highly enriched uranium) fuel elements, the proliferation problems related to HEU fuel and the generated high-level radioactive wastes, possible safety hazards of an interim storage of HEU containing wastes, for instance in the interim storage facility Ahaus, possible safety hazards of final disposal of HEU containing radioactive wastes, possibilities to avoid the use of HEU fuel in order to prevent further production of these wastes, requirement of processing spent HEU containing fuel elements for final disposal.

  2. The caCORE Software Development Kit: Streamlining construction of interoperable biomedical information services

    Directory of Open Access Journals (Sweden)

    Warzel Denise

    2006-01-01

    Full Text Available Abstract Background Robust, programmatically accessible biomedical information services that syntactically and semantically interoperate with other resources are challenging to construct. Such systems require the adoption of common information models, data representations and terminology standards as well as documented application programming interfaces (APIs. The National Cancer Institute (NCI developed the cancer common ontologic representation environment (caCORE to provide the infrastructure necessary to achieve interoperability across the systems it develops or sponsors. The caCORE Software Development Kit (SDK was designed to provide developers both within and outside the NCI with the tools needed to construct such interoperable software systems. Results The caCORE SDK requires a Unified Modeling Language (UML tool to begin the development workflow with the construction of a domain information model in the form of a UML Class Diagram. Models are annotated with concepts and definitions from a description logic terminology source using the Semantic Connector component. The annotated model is registered in the Cancer Data Standards Repository (caDSR using the UML Loader component. System software is automatically generated using the Codegen component, which produces middleware that runs on an application server. The caCORE SDK was initially tested and validated using a seven-class UML model, and has been used to generate the caCORE production system, which includes models with dozens of classes. The deployed system supports access through object-oriented APIs with consistent syntax for retrieval of any type of data object across all classes in the original UML model. The caCORE SDK is currently being used by several development teams, including by participants in the cancer biomedical informatics grid (caBIG program, to create compatible data services. caBIG compatibility standards are based upon caCORE resources, and thus the caCORE SDK has

  3. TRIGA Research Reactor Conversion to LEU and Modernization of Safety Related Systems

    Energy Technology Data Exchange (ETDEWEB)

    Sanda, R. M. [Institute for Nuclear Research Piteşti (SCN-Piteşti), Piteşti (Romania)

    2014-08-15

    The USA and IAEA proposed an international programme to reduce the enrichment of uranium in research reactors by converting nuclear fuel containing HEU into fuel containing 20% enriched uranium. The Government of Romania joined the programme and actively supported political, scientific, technical and economic actions that led to the conversion of the active area of the 14 MW TRIGA reactor at the Institute for Nuclear Research in Piteşti in May 2006. This confirmed the continuity of the Romanian Government’s non-proliferation policy and their active support of international cooperation. Conversion of the Piteşti research reactor was made possible by completion of milestones in the Research Agreement for Reactor Conversion, a contract signed with the US Department of Energy and Argonne National Laboratory. This agreement provided scientific and technical support and the possibility of delivery of all HEU TRIGA fuel to the United States. Additionally, about 65% of the fresh LEU fuel needed to start the conversion was delivered in the period 1992–1994. Furthermore, conversion was promoted through IAEA Technical Cooperation project ROM/4/024 project funded primarily by the United States that supported technical and scientific efforts and the delivery of the remaining required LEU nuclear fuel to complete the conversion. Nuclear fuel to complete the conversion was made by the French company CERCA with a tripartite contract among the IAEA, CERCA and Romania. The contract was funded by the US Department of Energy with a voluntary contribution by the Romanian Government. The contract stipulated manufacturing and delivery of LEU fuel by CERCA with compliance measures for quality, delivery schedule and safety requirements set by IAEA standards and Romanian legislation. The project was supported by the ongoing technical cooperation, safeguards, legal and procurement assistance of the IAEA, in particular its Department of Nuclear Safety. For Romanian research, the

  4. The seismic assessment of radially keyed graphite moderator cores

    International Nuclear Information System (INIS)

    Steer, A.G.; Payne, J.F.B.

    1996-01-01

    The modelling of AGR and Magnox cores has to deal with the very large number of components that make up the core, and the non-linear response due to the clearances in the keying system. This paper examines the conditions under which it is permissible to linearise the response. By comparing the results of discrete and continuum models of the core, the paper also shows that the number of components in the core is sufficiently large that the core can be approximated satisfactorily by an anisotropic solid material. The material has unusual properties, but these can be handled within the standard framework for the description of the elastic properties of an anisotropic solid. This description of the core by an equivalent solid material can readily be incorporated into finite element models of the reactor internal structure. Such models have been set up for both AGR and Magnox reactors. The models are being used to assess the seismic response of these reactors. (author). 5 refs, 6 figs

  5. Study on HANARO core conversion using U-Mo fuel

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Seo, C.G.; Park, S.J.; Kim, H.; Kim, C.K.

    2002-01-01

    Two types of fuel rods with different fuel meat diameter and uranium density are considered for HANARO core conversion with high density U-Mo fuel. Arranging standard fuels of 5.0 g U/cc and 6.35 mm in diameter at the inner ring of an assembly and reduced fuels of 4.3 g U/cc and 5.49 mm in diameter at the outer ring of an assembly flattens the assembly power distribution and avoids the increase of linear heat generation rate due to using higher uranium density and less number of fuel rods. The maximum linear heat generation rate is similar with the current reference core and four fuel sites at the outer core in the reflector tank is converted to the irradiation sites to suit more demand on fuel tests and radioisotope production at outer core sites. This new core has 32% longer fuel cycle than the current reference core. (author)

  6. Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

    Science.gov (United States)

    Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

    2010-10-01

    The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

  7. Analysis of excess reactivity of JOYO MK-III performance test core

    International Nuclear Information System (INIS)

    Maeda, Shigetaka; Yokoyama, Kenji

    2003-10-01

    JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

  8. Standards for vision science libraries: 2014 revision.

    Science.gov (United States)

    Motte, Kristin; Caldwell, C Brooke; Lamson, Karen S; Ferimer, Suzanne; Nims, J Chris

    2014-10-01

    This Association of Vision Science Librarians revision of the "Standards for Vision Science Libraries" aspires to provide benchmarks to address the needs for the services and resources of modern vision science libraries (academic, medical or hospital, pharmaceutical, and so on), which share a core mission, are varied by type, and are located throughout the world. Through multiple meeting discussions, member surveys, and a collaborative revision process, the standards have been updated for the first time in over a decade. While the range of types of libraries supporting vision science services, education, and research is wide, all libraries, regardless of type, share core attributes, which the standards address. The current standards can and should be used to help develop new vision science libraries or to expand the growth of existing libraries, as well as to support vision science librarians in their work to better provide services and resources to their respective users.

  9. Eliminating Stockpiles of Highly Enriched Uranium. Options for an Action Agenda in Co-operation with the Russian Federation. Report submitted to the Swedish Ministry for Foreign Affairs

    International Nuclear Information System (INIS)

    Arbman, Gunnar; Calogero, Francesco; Martellini, Maurizio; Bremer Maerli, Morten; Nikitin, Alexander; Prawitz, Jan

    2004-04-01

    This study is of an exploratory nature. It provides preliminary assessments of issues of relevance for HEU elimination in Russia including: (a) technical issues concerning the HEU down-blending; uranium transparency and verification requirements; description of current Russian HEU locations; the HEU down-blending capacities, and the HEU logistics, and (b) various political and financial requirements and considerations. For future, practical project measures to be put in place, further investigations that deal with HEU logistics and handling are needed. Such studies - that obviously should include and engage key Russian actors - are possible, if they take legitimate Russian security and sensitivity concerns into consideration. Interestingly, there is a growing perception in Russia that large stocks of HEU are not required and that they could, in fact, constitute a source of danger

  10. Young Adult Literature and the Common Core: A Surprisingly Good Fit

    Science.gov (United States)

    Ostenson, Jonathan; Wadham, Rachel

    2012-01-01

    Advocates have long argued that an increased role for young adult literature in the classroom would help students' reading development. At first glance, the widely adopted Common Core State Standards might seem in opposition to an increased role for such literature. A closer examination of the common core documents suggests, however, that young…

  11. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  12. Core Competencies in Advanced Training: What Supervisors Say about Graduate Training

    Science.gov (United States)

    Nelson, Thorana S.; Graves, Todd

    2011-01-01

    In an attempt to identify needed mental health skills, many professional organizations have or are in the process of establishing core competency standards for their professions. The AAMFT identified 128 core competencies for the independent practice of MFT. The aim of this study was to learn the opinions of AAMFT Approved Supervisors as to how…

  13. Low enrichment fuel conversion for Iowa State University

    International Nuclear Information System (INIS)

    Rohach, A.F.; Hendrickson, R.A.

    1991-08-01

    Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. This included development of procedures and tools for the disassembly process. During the period we held many practice sessions applying these tools and practices to a dummy fuel assembly. The LEU fuel was received on April 10, 1991 and the reactor was shut down on May 3, 1991 for refueling. The twelve HEU fuel assemblies in the UTR-10 reactor core were removed and disassembled during the week of May 6--9, 1991. The disassembly process went smoothly with only a few minor problems. Also during this reporting period several experimental measurements and preventative maintenance tasks were accomplished. Finally procedures and practices have been developed for the new LEU fuel loading and critical experiments which are to be completed during the late summer of 1991

  14. Fracture resistance of upper central incisors restored with different posts and cores

    Directory of Open Access Journals (Sweden)

    Maryam Rezaei Dastjerdi

    2015-08-01

    Full Text Available Objectives To determine and compare the fracture resistance of endodontically treated maxillary central incisors restored with different posts and cores. Materials and Methods Forty-eight upper central incisors were randomly divided into four groups: cast post and core (group 1, fiber-reinforced composite (FRC post and composite core (group 2, composite post and core (group 3, and controls (group 4. Mesio-distal and bucco-lingual dimensions at 7 and 14 mm from the apex were compared to ensure standardization among the groups. Twelve teeth were prepared for crown restoration (group 4. Teeth in other groups were endodontically treated, decoronated at 14 mm from the apex, and prepared for posts and cores. Resin-based materials were used for cementation in groups 1 and 2. In group 3, composite was used directly to fill the post space and for core build-up. All samples were restored by standard metal crowns using glass ionomer cement, mounted at 135° vertical angle, subjected to thermomechanical aging, and then fractured using a universal testing machine. Kruskal-Wallis and Mann-Whitney U tests were used to analyze the data. Results Fracture resistance of the groups was as follows: Control (group 4 > cast post and core (group 1 > fiber post and composite core (group 2 > composite post and core (group 3. All samples in groups 2 and 3 fractured in restorable patterns, whereas most (58% in group 1 were non-restorable. Conclusions Within the limitations of this study, FRC posts showed acceptable fracture resistance with favorable fracture patterns for reconstruction of upper central incisors.

  15. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  16. Mapping of a standard documentation template to the ICF core sets for arthritis and low back pain.

    Science.gov (United States)

    Escorpizo, Reuben; Davis, Kandace; Stumbo, Teri

    2010-12-01

    To identify the contents of a documentation template in The Guide to Physical Therapist Practice using the International Classification of Functioning, Disability, and Health (ICF) Core Sets for rheumatoid arthritis, osteoarthritis, and low back pain (LBP) as reference. Concepts were identified from items of an outpatient documentation template and mapped to the ICF using established linking rules. The ICF categories that were linked were compared with existing arthritis and LBP Core Sets. Based on the ICF, the template had the highest number (29%) of linked categories under Activities and participation while Body structures had the least (17%). ICF categories in the arthritis and LBP Core Sets had a 37-55% match with the ICF categories found in the template. We found 164 concepts that were not classified or not defined and 37 as personal factors. The arthritis and LBP Core Sets were reflected in the contents of the template. ICF categories in the Core Sets were reflected in the template (demonstrating up to 55% match). Potential integration of ICF in documentation templates could be explored and examined in the future to enhance clinical encounters and multidisciplinary communication. Copyright © 2010 John Wiley & Sons, Ltd.

  17. Reduced Enrichment for Research and Test Reactors. Proceedings of the XVIII international meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    Almost 50 papers presented were showing the status of the national programs related to conversion of research reactor cores from highly enriched (HEU) to low enriched uranium (LEU) fuel elements. Design of new fuel elements (uranium silicides) and safety related calculations were dealt with taking into account fuel cycle issues, meaning spent fuel storage and transportation. A number of presentations were devoted to Mo-99 production using LEU targets.

  18. Reduced Enrichment for Research and Test Reactors. Proceedings of the XVIII international meeting

    International Nuclear Information System (INIS)

    2004-01-01

    Almost 50 papers presented were showing the status of the national programs related to conversion of research reactor cores from highly enriched (HEU) to low enriched uranium (LEU) fuel elements. Design of new fuel elements (uranium silicides) and safety related calculations were dealt with taking into account fuel cycle issues, meaning spent fuel storage and transportation. A number of presentations were devoted to Mo-99 production using LEU targets

  19. Abstracts and Papers of the 2003 International RERTR Meeting

    International Nuclear Information System (INIS)

    2003-10-01

    The papers presented at the 25th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) were classified according to the following subjects: National programs covering shipment of HEU to countries of origin and national RERTR programs; Molybdenum 99 production by using LEU fuel; Development and fabrication of new type of LEU fuel elements; Reactor core conversion analysis; Safety, spent fuel and regulatory issues; spent fuel storage and management

  20. Taking a Comprehensive Approach to Common Core Rollout

    Science.gov (United States)

    Illingworth, Amy

    2016-01-01

    One district, South Bay Union School District, devises and executes a comprehensive strategy that includes training for district staff, teachers and coaches, including using PLCs and professional development in the implementation of Common Core English language arts standards.

  1. Core to College Evaluation: Statewide Networks. Connecting Education Systems and Stakeholders to Support College Readiness

    Science.gov (United States)

    Bracco, Kathy Reeves; Klarin, Becca; Broek, Marie; Austin, Kim; Finkelstein, Neal; Bugler, Daniel; Mundry, Susan

    2014-01-01

    The Core to College initiative aims to facilitate greater coordination between K-12 and postsecondary education systems around implementation of the Common Core State Standards and aligned assessments. Core to College grants have been awarded to teams in Colorado, Florida, Hawaii, Kentucky, Louisiana, Massachusetts, North Carolina, Oregon,…

  2. The impact of the EUSCLE Core Set Questionnaire for the assessment of cutaneous lupus erythematosus.

    Science.gov (United States)

    Kuhn, A; Patsinakidis, N; Bonsmann, G

    2010-08-01

    Epidemiological data and standard European guidelines for the diagnosis and treatment of cutaneous lupus erythematosus (CLE) are lacking in the current literature. In order to provide a standardized tool for an extensive consistent data collection, a study group of the European Society of Cutaneous Lupus Erythematosus (EUSCLE) recently developed a Core Set Questionnaire for the assessment of patients with different subtypes of CLE. The EUSCLE Core Set Questionnaire includes six sections on patient data, diagnosis, skin involvement, activity and damage of disease, laboratory analysis, and treatment. An instrument like the EUSCLE Core Set Questionnaire is essential to gain a broad and comparable data collection of patients with CLE from different European centres and to achieve consensus concerning clinical standards for the disease. The data will also be important for further characterization of the different CLE subtypes and the evaluation of therapeutic strategies; moreover, the EUSCLE Core Set Questionnaire might also be useful for the comparison of data in clinical trials. In this review, the impact of the EUSCLE Core Set Questionnaire is discussed in detail with regard to clinical and serological features as well as therapeutic modalities in CLE.

  3. Kenaf Core Particleboard and Its Sound Absorbing Properties

    OpenAIRE

    Mohamad Jani Saad; Izran Kamal

    2012-01-01

    In this study, kenaf (Hibiscus cannabinus) core particleboards as insulation boards were manufactured. The boards were fabricated with three different densities i.e. 350 kg/m3, 450 kg/m3 and 550 kg/m3 at urea formaldehyde resin (UF) loadings of 8%, 10% and 12% (w/w) based on the dry weight of the kenaf core particles. The fabricated boards were evaluated for its noise acoustical coefficients (NAC) by following the ASTM E1050-98 standard requirements. The study revealed that boards with higher...

  4. Static analysis of material testing reactor cores:critical core calculations

    International Nuclear Information System (INIS)

    Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

    1999-01-01

    A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

  5. Ultra-Performance Liquid Chromatography Hyphenated with Quadrupole-Orbitrap Mass Spectrometry for Simultaneous Determination of Necine-Core-Structure Pyrrolizidine Alkaloids in Crotalaria sessiliflora L. without all Corresponding Standards.

    Science.gov (United States)

    Zhang, Wei; Huai, Wenbei; Zhang, Yi; Shen, Jincan; Tang, Xunyou; Xie, Xiujuan; Wang, Ke; Fan, Huajun

    2017-09-01

    Crotalaria sessiliflora L. is a Chinese traditional herb for treatment of cutaneum carcinoma and cervical carcinoma. In addition to monocrotaline, coexisting pyrrolizidine alkaloids (PAs) also require further quantification for quality control and pharmaceutical uses of the herb. To establish a UPLC-Q-Orbitrap/MS method of simultaneous determination of coexisting PAs with same parent structure for quality control and comprehensive researches of Crotalaria sessiliflora L. PAs in Crotalaria sessiliflora L. were analysed by UPLC-Q-Orbitrap/MS method. Coexisting PAs were identified by mass data of full MS-dd-MS 2 based on the characteristic fragmentation pattern and necine-core structure. Moreover, quantification of PAs was conducted by parallel reaction monitoring (PRM) mode using m/z 138, m/z 120 and m/z 94 from identical necine-core structure as quantitative ions with single monocrotaline standard for accurate calibration. Five PAs, named monocrotaline, retrorsine, senecionine, integerrimine, O-9-angeloylretronecine, were indentified and confirmed. Quantitative ions of m/z 138, m/z 120 and m/z 94 were used for quantification of PAs containing the necine-core structure in Crotalaria sessiliflora L. The results demonstrated that contents, precision and recoveries of the five PAs mentioned earlier were respectively 3.307-30.35 μg/g, 1.1-4.5% and 88.91-92.33% while using m/z 120 as the best quantitative ion. The UPLC-Q-Orbitrap/MS method was established for simultaneous determination of five PAs in Crotalaria sessiliflora L. without all corresponding standards, and was proved that it was simple, convenient and effective for comprehensive quality control and pharmaceutical uses. Copyright © 2017 John Wiley & Sons, Ltd. Copyright © 2017 John Wiley & Sons, Ltd.

  6. Innovative nuclear thermal rocket concept utilizing LEU fuel for space application

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Venneri, Paolo; Choi, Jae Young; Jeong, Yong Hoon; Chang, Soon Heung

    2015-01-01

    Space is one of the best places for humanity to turn to keep learning and exploiting. A Nuclear Thermal Rocket (NTR) is a viable and more efficient option for human space exploration than the existing Chemical Rockets (CRs) which are highly inefficient for long-term manned missions such as to Mars and its satellites. NERVA derived NTR engines have been studied for the human missions as a mainstream in the United States of America (USA). Actually, the NERVA technology has already been developed and successfully tested since 1950s. The state-of-the-art technology is based on a Hydrogen gas (H_2) cooled high temperature reactor with solid core utilizing High-Enriched Uranium (HEU) fuel to reduce heavy metal mass and to use fast or epithermal neutron spectrums enabling simple core designs. However, even though the NTR designs utilizing HEU is the best option in terms of rocket performance, they inevitably provoke nuclear proliferation obstacles on all Research and Development (R and D) activities by civilians and non-nuclear weapon states, and its eventual commercialization. To surmount the security issue to use HEU fuel for a NTR, a concept of the innovative NTR engine, Korea Advanced NUclear Thermal Engine Rocket utilizing Low-Enriched Uranium fuel (KANUTER-LEU) is presented in this paper. The design goal of KANUTER-LEU is to make use of a LEU fuel for its compact reactor, but does not sacrifice the rocket performance relative to the traditional NTRs utilizing HEU. KANUTER-LEU mainly consists of a fission reactor utilizing H_2 propellant, a propulsion system and an optional Electricity Generation System as a bimodal engine. To implement LEU fuel for the reactor, the innovative engine adopts W-UO_2 CERMET fuel to drastically increase uranium density and thermal neutron spectrum to improve neutron economy in the core. The moderator and structural material selections also consider neutronic and thermo-physical characteristics to reduce non-fission neutron loss and

  7. The Yucca Mountain Project prototype air-coring test, U12g tunnel, Nevada test site

    International Nuclear Information System (INIS)

    Ray, J.M.; Newsom, J.C.

    1994-12-01

    The Prototype Air-Coring Test was conducted at the Nevada Test Site (NTS) G-Tunnel facility to evaluate standard coring techniques, modified slightly for air circulation, for use in testing at a prospective nuclear waste repository at Yucca Mountain, Nevada. Air-coring technology allows sampling of subsurface lithology with minimal perturbation to ambient characteristic such as that required for exploratory holes near aquifers, environmental applications, and site characterization work. Two horizontal holes were cored, one 50 ft long and the other 150 ft long, in densely welded fractured tuff to simulate the difficult drilling conditions anticipated at Yucca Mountain. Drilling data from seven holes on three other prototype tests in nonwelded tuff were also collected for comparison. The test was used to establish preliminary standards of performance for drilling and dust collection equipment and to assess procedural efficiencies. The Longyear-38 drill achieved 97% recovery for HQ-size core (-2.5 in.), and the Atlas Copco dust collector (DCT-90) captured 1500 lb of fugitive dust in a mine environment with only minor modifications. Average hole production rates were 6-8 ft per 6-h shift in welded tuff and almost 20 ft per shift on deeper holes in nonwelded tuff. Lexan liners were successfully used to encapsulate core samples during the coring process and protect core properties effectively. The Prototype Air-Coring Test demonstrated that horizontal air coring in fractured welded tuff (to at least 150 ft) can be safely accomplished by proper selection, integration, and minor modification of standard drilling equipment, using appropriate procedures and engineering controls. The test also indicated that rig logistics, equipment, and methods need improvement before attempting a large-scale dry drilling program at Yucca Mountain

  8. Lesson Planning with the Common Core

    Science.gov (United States)

    Estes, Linda A.; McDuffie, Amy Roth; Tate, Cathie

    2014-01-01

    Planning a lesson can be similar to planning a road trip--a metaphor the authors use to describe how they applied research and theory to their lesson planning process. A map and mode of transportation, the Common Core State Standards for Mathematics (CCSSM) and textbooks as resources, can lead to desired destinations, such as students engaging in…

  9. Accessing the Common Core Standards for Students with Learning Disabilities: Strategies for Writing Standards-Based IEP Goals

    Science.gov (United States)

    Caruana, Vicki

    2015-01-01

    Since the reauthorization of the Individuals With Disabilities Education Act (IDEA) in 2004, standards-based individualized education plans (IEPs) have been an expectation for serving students with disabilities in the K-12 public school setting. Nearly a decade after the mandates calling for standards-based IEPs, special educators still struggle…

  10. Development of new core competencies for Taiwanese Emergency Medical Technicians

    Directory of Open Access Journals (Sweden)

    Chang YT

    2018-03-01

    Full Text Available Yu-Tung Chang,1,2 Kuang-Chau Tsai,2 Brett Williams1,3 1Department of Community Emergency Health and Paramedic Practice, Faculty of Medicine, Nursing and Health Sciences, Monash University, Frankston, VIC, Australia; 2Department of Emergency Medicine, Far Eastern Memorial Hospital, New Taipei City, Taiwan; 3Division of Paramedicine, University of Tasmania, Hobart, TAS, Australia Objectives: Core competencies are considered the foundation for establishing Emergency Medical Technician (EMT and paramedic curricula, and for ensuring performance standards in the delivery of prehospital care. This study surveyed EMT instructors and medical directors to identify the most desirable core competencies for all levels of EMTs in Taiwan. Methods: A principal components analysis with Varimax rotation was conducted. An online questionnaire was distributed to obtain perspectives of EMT instructors and medical directors on the most desirable core competencies for EMTs. The target population was EMT training-course instructors and medical directors of fire departments in Taiwan. The questionnaire comprised 61 competency items, and multiple-choice and open-ended questions were used to obtain respondents’ perspectives of the Taiwanese EMT training and education system. Results: The results identified three factors at EMT-1 and EMT-2 levels and five factors at the EMT-Paramedic level. The factors for EMT-1 and EMT-2 were similar, and those for EMT-Paramedics identified further comprehensive competence perspectives. The key factors that appear to influence the development of the Taiwanese Emergency Medical Services (EMS education system are the attitude of authorities, the licensure system, and legislation. Conclusion: The findings present new core competencies for the Taiwanese EMT system and provide capacity to redesign curricula and reconsider roles for EMT-1 and EMT-2 technicians. At the EMT-Paramedic level, the findings demonstrate the importance of

  11. Thermal-hydraulic characteristics of double flat core HCLWR

    International Nuclear Information System (INIS)

    Sugimoto, Jun; Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio

    1989-02-01

    A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the same time, high conversion ratio and high burnup can be obtainable. The proposed double flat core HCLWR, based on these physical advantages and the consideration of safety assurance, aims at efficient use of the pressure vessel space to produce comparable thermal output as current 3-loop PWRs. The present work revealed the following items concerning the thermalhydraulic feasibility of the double flat core HCLWR: (1) Main thermal-hydraulic parameters of the plant can be almost the same as current PWRs, showing the use of PWR standard components without major modifications except in core region. (2) Heat removal from the fuel rod in a steady operational condition has enough margin to the critical heat flux (CHF) limit, which is evaluated with the existing CHF correlations. (3) The calculation by REFLA code shows that the maximum cladding temperature in LOCA-reflood is estimated to be far lower than the licensing criteria. It is therefore considered that the proposed double flat core HCLWR is feasible from the point of thermal-hydraulics. Since the available data base has certain applicational limit to the very short core as the present double flat core HCLWR, further detailed assessment is required. (author)

  12. WIMS/PANTHER analysis of UO2/MOX cores using embedded super-cells

    International Nuclear Information System (INIS)

    Knight, M.; Bryce, P.; Hall, S.

    2012-01-01

    This paper describes a method of analysing PWR UO 2 MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

  13. Designing Class Activities to Meet Specific Core Training Competencies: A Developmental Approach

    Science.gov (United States)

    Guth, Lorraine J.; McDonnell, Kelly A.

    2004-01-01

    This article presents a developmental model for designing and utilizing class activities to meet specific Association for Specialists in Group Work (ASGW) core training competencies for group workers. A review of the relevant literature about teaching group work and meeting core training standards is provided. The authors suggest a process by…

  14. The Immune System of HIV-Exposed Uninfected Infants.

    Science.gov (United States)

    Abu-Raya, Bahaa; Kollmann, Tobias R; Marchant, Arnaud; MacGillivray, Duncan M

    2016-01-01

    Infants born to human immunodeficiency virus (HIV) infected women are HIV-exposed but the majority remains uninfected [i.e., HIV-exposed uninfected (HEU)]. HEU infants suffer greater morbidity and mortality from infections compared to HIV-unexposed (HU) peers. The reason(s) for these worse outcomes are uncertain, but could be related to an altered immune system state. This review comprehensively summarizes the current literature investigating the adaptive and innate immune system of HEU infants. HEU infants have altered cell-mediated immunity, including impaired T-cell maturation with documented hypo- as well as hyper-responsiveness to T-cell activation. And although prevaccination vaccine-specific antibody levels are often lower in HEU than HU, most HEU infants mount adequate humoral immune response following primary vaccination with diphtheria toxoid, haemophilus influenzae type b, whole cell pertussis, measles, hepatitis B, tetanus toxoid, and pneumococcal conjugate vaccines. However, HEU infants are often found to have lower absolute neutrophil counts as compared to HU infants. On the other hand, an increase of innate immune cytokine production and expression of co-stimulatory markers has been noted in HEU infants, but this increase appears to be restricted to the first few weeks of life. The immune system of HEU children beyond infancy remains largely unexplored.

  15. Design of a reactor core in the Oma Full MOX-ABWR

    International Nuclear Information System (INIS)

    Hama, Teruo

    1999-01-01

    The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

  16. Improvement of Cycle Dependent Core Model for NPP Simulator

    International Nuclear Information System (INIS)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-01

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations

  17. Improvement of Cycle Dependent Core Model for NPP Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Song, J. S.; Koo, B. S.; Kim, H. Y. and others

    2003-11-15

    The purpose of this study is to establish automatic core model generation system and to develop 4 cycle real time core analysis methodology with 5% power distribution and 500 pcm reactivity difference criteria for nuclear power plant simulator. The standardized procedure to generate database from ROCS and ANC, which are used for domestic PWR core design, was established for the cycle specific simulator core model generation. An automatic data interface system to generate core model also established. The system includes ARCADIS which edits group constant and DHCGEN which generates interface coupling coefficient correction database. The interface coupling coefficient correction method developed in this study has 4 cycle real time capability and accuracies of which the maximum differences between core design results are within 103 pcm reactivity, 1% relative power distribution and 6% control rod worth. A nuclear power plant core simulation program R-MASTER was developed using the methodology and applied by the concept of distributed client system in simulator. The performance was verified by site acceptance test in Simulator no. 2 in Kori Training Center for 30 initial condition generation and 27 steady state, transient and postulated accident situations.

  18. Convective heating of the inner core of red giants prior to the peak of the core helium flash

    International Nuclear Information System (INIS)

    Cole, P.W.; Demarque, P.; Deupree, R.G.

    1985-01-01

    The effects of convective overshooting across the temperature inversion in the cores of red giants are investigated from the onset of the core convection zone to the peak of the core helium flash using a model for overshooting in stellar evolution, based on two-dimensional and three-dimensional hydrodynamic simulations of the core helium flash. A major effect of the overshooting is the substantial heating of the material interior to the temperature inversion, producing a smoother temperature profile. This interior heating is thus unimportant until approximately 1 week preceding the time of maximum temperature, but then produces temperature changes on a time scale short with respect to the evolution time scale. Interior heating (1) alters the standard relation of the maximum temperature and the density at the point of maximum temperature, (2) makes the maximum temperature occur at a smaller mass fraction, (3) causes the time of maximum temperature to occur hundreds of years earlier in the red giant evolution, and (4) redistributes the mass from the location of maximum temperature. Since the degree of degeneracy is known to affect the violence of the flash in the hydrodynamic phase, internal heating may play an important role in determining the subsequent evolution of the core

  19. Cusp-core problem and strong gravitational lensing

    International Nuclear Information System (INIS)

    Li Nan; Chen Daming

    2009-01-01

    Cosmological numerical simulations of galaxy formation have led to the cuspy density profile of a pure cold dark matter halo toward the center, which is in sharp contradiction with the observations of the rotation curves of cold dark matter-dominated dwarf and low surface brightness disk galaxies, with the latter tending to favor mass profiles with a flat central core. Many efforts have been devoted to resolving this cusp-core problem in recent years, among them, baryon-cold dark matter interactions are considered to be the main physical mechanisms erasing the cold dark matter (CDM) cusp into a flat core in the centers of all CDM halos. Clearly, baryon-cold dark matter interactions are not customized only for CDM-dominated disk galaxies, but for all types, including giant ellipticals. We first fit the most recent high resolution observations of rotation curves with the Burkert profile, then use the constrained core size-halo mass relation to calculate the lensing frequency, and compare the predicted results with strong lensing observations. Unfortunately, it turns out that the core size constrained from rotation curves of disk galaxies cannot be extrapolated to giant ellipticals. We conclude that, in the standard cosmological paradigm, baryon-cold dark matter interactions are not universal mechanisms for galaxy formation, and therefore, they cannot be true solutions to the cusp-core problem.

  20. Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Joo S.; Diamond, David

    2016-12-06

    A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

  1. Building the Missing Link between the Common Core and Improved Learning

    Science.gov (United States)

    Rodde, Amy Coe; McHugh, Lija

    2013-01-01

    The Common Core State Standards, adopted by 45 states and the District of Columbia, raise the bar for what students need to learn at each stage of their K-12 education. The goal is to better prepare students for college and careers. The most important thing that education leaders can do to help the Common Core succeed is to support teachers in…

  2. Conversion (Utilizing LEU instead HEU) of research reactors in Czech Republic

    International Nuclear Information System (INIS)

    Matejka, K.; Sklenka, L.; Listik, E.; Ernest, J.

    1998-01-01

    This paper shortly describes some advantages on the RERTR-programme in the Czech Republic. Further calculations and experimental measurements finished on the VR-1 training reactor Sparrow. Paper brings results and its evaluation as well as one-year operation experiences with the Russian fuel assemblies IRT-3M and also operation experiments with mixed Core configuration (the Russian fuel assemblies IRT-2M with enrichment 80% 235 U and 36% 235 U) on the LVR-15 research reactor. (author)

  3. Core strength training for patients with chronic low back pain.

    Science.gov (United States)

    Chang, Wen-Dien; Lin, Hung-Yu; Lai, Ping-Tung

    2015-03-01

    [Purpose] Through core strength training, patients with chronic low back pain can strengthen their deep trunk muscles. However, independent training remains challenging, despite the existence of numerous core strength training strategies. Currently, no standardized system has been established analyzing and comparing the results of core strength training and typical resistance training. Therefore, we conducted a systematic review of the results of previous studies to explore the effectiveness of various core strength training strategies for patients with chronic low back pain. [Methods] We searched for relevant studies using electronic databases. Subsequently, we evaluated their quality by analyzing the reported data. [Results] We compared four methods of evaluating core strength training: trunk balance, stabilization, segmental stabilization, and motor control exercises. According to the results of various scales and evaluation instruments, core strength training is more effective than typical resistance training for alleviating chronic low back pain. [Conclusion] All of the core strength training strategies examined in this study assist in the alleviation of chronic low back pain; however, we recommend focusing on training the deep trunk muscles to alleviate chronic low back pain.

  4. Assisting Pupils in Mathematics Achievement (The Common Core Standards)

    Science.gov (United States)

    Ediger, Marlow

    2011-01-01

    Mathematics teachers must expect reasonably high standards of achievement from pupils. Too frequently, pupils attain at a substandard level and more optimal achievement is necessary. Thus, pupils should have self esteem needs met in the school and classroom setting. Thus, learners feel that mathematics is worthwhile and effort must be put forth to…

  5. Exploring Function Transformations Using the Common Core

    Science.gov (United States)

    Hall, Becky; Giacin, Rich

    2013-01-01

    When examining transformations of the plane in geometry, teachers typically have students experiment with transformations of polygons. Students are usually quick to notice patterns with ordered pairs. The Common Core State Standard, Geometry, Congruence 2 (G-CO.2), requires students to describe transformations as functions that take points in the…

  6. Understanding core conductor fabrics

    International Nuclear Information System (INIS)

    Swenson, D E

    2011-01-01

    ESD Association standard test method ANSI/ESD STM2.1 - Garments (STM2.1), provides electrical resistance test procedures that are applicable for materials and garments that have surface conductive or surface dissipative properties. As has been reported in other papers over the past several years 1 fabrics are now used in many industries for electrostatic control purposes that do not have surface conductive properties and therefore cannot be evaluated using the procedures in STM2.1 2 . A study was conducted to compare surface conductive fabrics with samples of core conductor fibre based fabrics in order to determine differences and similarities with regards to various electrostatic properties. This work will be used to establish a new work item proposal within WG-2, Garments, in the ESD Association Standards Committee in the USA.

  7. Core Hunter 3: flexible core subset selection.

    Science.gov (United States)

    De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

    2018-05-31

    Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

  8. An approach to development of structural design criteria for highly irradiated core components

    International Nuclear Information System (INIS)

    Nelson, D.V.

    1980-01-01

    The advent of the fast breeder reactor presents novel challenges in structural design and materials engineering. For instance, the core components of these reactors experience high energy neutron irradiation at elevated temperature, which causes significant time-dependent changes in material behaviour, such as a progressive loss of ductility. New structural design criteria are needed to extend elevated temperature design-by-analysis to account for these changes. Alloys best able to cope with the demands of the core operating environment are being explored and their structural behaviour characterized. The purpose of this paper is to illustrate an approach used in the development of core component structural design criteria. To do this, several design rules, plus brief rationale, from draft RDT Standards F9-7, -8 and -9 will be presented. These recently completed standards ('Structural Design Guidelines for Breeder Reactor Core Components') were prepared for the U.S. Department of Energy and represent a consensus among most organizations participating in the U.S. breeder program. (author)

  9. Telemetric measurement of body core temperature in exercising unconditioned Labrador retrievers.

    Science.gov (United States)

    Angle, T Craig; Gillette, Robert L

    2011-04-01

    This project evaluated the use of an ingestible temperature sensor to measure body core temperature (Tc) in exercising dogs. Twenty-five healthy, unconditioned Labrador retrievers participated in an outdoor 3.5-km run, completed in 20 min on a level, 400-m grass track. Core temperature was measured continuously with a telemetric monitoring system before, during, and after the run. Data were successfully collected with no missing data points during the exercise. Core temperature elevated in the dogs from 38.7 ± 0.3°C at pre-exercise to 40.4 ± 0.6°C post-exercise. While rectal temperatures are still the standard of measurement, telemetric core temperature monitors may offer an easier and more comfortable means of sampling core temperature with minimal human and mechanical interference with the exercising dog.

  10. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  11. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  12. MA-core loaded untuned RF compression cavity for HIRFL-CSR

    International Nuclear Information System (INIS)

    Mei Lirong; Xu Zhe; Yuan Youjin; Jin Peng; Bian Zhibin; Zhao Hongwei; Xia Jiawen

    2012-01-01

    To meet the requirements of high energy density physics and plasma physics research at HIRFL-CSR the goal of achieving a higher accelerating gap voltage was proposed. Therefore, a magnetic alloy (MA)-core loaded radio frequency (RF) cavity that can provide a higher accelerating gap voltage compared to standard ferrite loaded cavities has been studied at IMP. In order to select the proper magnetic alloy material to load the RF compression cavity, measurements of four different kinds of sample MA-cores have been carried out. By testing the small cores, the core composition was selected to obtain the desired performance. According to the theoretical calculation and simulation, which show reasonable consistency for the MA-core loaded cavity, the desired performance can be achieved. Finally about 1000 kW power will be needed to meet the requirements of 50 kV accelerating gap voltage by calculation.

  13. Valence-to-core-detected X-ray absorption spectroscopy

    DEFF Research Database (Denmark)

    Hall, Eleanor R.; Pollock, Christopher J.; Bendix, Jesper

    2014-01-01

    X-ray absorption spectroscopy (XAS) can provide detailed insight into the electronic and geometric structures of transition-metal active sites in metalloproteins and chemical catalysts. However, standard XAS spectra inherently represent an average contribution from the entire coordination...... environment with limited ligand selectivity. To address this limitation, we have investigated the enhancement of XAS features using valence-to-core (VtC)-detected XAS, whereby XAS spectra are measured by monitoring fluorescence from valence-to-core X-ray emission (VtC XES) events. VtC emission corresponds...... to transitions from filled ligand orbitals to the metal 1s core hole, with distinct energetic shifts for ligands of differing ionization potentials. VtC-detected XAS data were obtained from multiple valence emission features for a series of well-characterized Mn model compounds; taken together, these data...

  14. Design and experience of HEU and LEU fuel for WWR-M reactor

    International Nuclear Information System (INIS)

    Enin, A.A.; Erykalov, A.N.; Zakharov, A.S.; Zvezdkin, V.S.; Kirsanov, G.A.; Konoplev, K.A.; L'vov, V.S.; Petroc, Y.V.; Saikov, Y.P.

    1997-01-01

    A research reactor for providing high neutron fluxes has to have a compact, well breeding core with high specific heat removal. The WWR-M fuel elements meet these demands. They have optimum metal-to-water ratio and the recordly developed specific heat-transfer surface providing in a pool-type reactor at atmospheric pressure the unit heat of (900±100) kW. (author)

  15. WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells

    Energy Technology Data Exchange (ETDEWEB)

    Knight, M.; Bryce, P. [EDF Energy, Barnett Way, Barnwood, Gloucester (United Kingdom); Hall, S. [Advanced Modelling and Computation Group, Imperial College, London (United Kingdom)

    2012-07-01

    This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)

  16. Altered Natural Killer Cell Function in HIV-Exposed Uninfected Infants

    Directory of Open Access Journals (Sweden)

    Christiana Smith

    2017-04-01

    Full Text Available ObjectivesHIV-exposed uninfected (HEU infants have higher rates of severe and fatal infections compared with HIV-unexposed (HUU infants, likely due to immune perturbations. We hypothesized that alterations in natural killer (NK cell activity might occur in HEU infants and predispose them to severe infections.DesignCase–control study using cryopreserved peripheral blood mononuclear cells (PBMCs at birth and 6 months from HEU infants enrolled from 2002 to 2009 and HUU infants enrolled from 2011 to 2013.MethodsNK cell phenotype and function were assessed by flow cytometry after 20-h incubation with and without K562 cells.ResultsThe proportion of NK cells among PBMCs was lower at birth in 12 HEU vs. 22 HUU (1.68 vs. 10.30%, p < 0.0001 and at 6 months in 52 HEU vs. 72 HUU (3.09 vs. 4.65%, p = 0.0005. At birth, HEU NK cells demonstrated increased killing of K562 target cells (p < 0.0001 and increased expression of CD107a (21.65 vs. 12.70%, p = 0.047, but these differences resolved by 6 months. Stimulated HEU NK cells produced less interferon (IFNγ at birth (0.77 vs. 2.64%, p = 0.008 and at 6 months (4.12 vs. 8.39%, p = 0.001, and showed reduced perforin staining at 6 months (66.95 vs. 77.30%, p = 0.0008. Analysis of cell culture supernatants indicated that lower NK cell activity in HEU was associated with reduced interleukin (IL-12, IL-15, and IL-18. Addition of recombinant human IL-12 to stimulated HEU PBMCs restored IFNγ production to that seen in stimulated HUU cultures.ConclusionNK cell proportion, phenotype, and function are altered in HEU infants. NK cell cytotoxicity and degranulation are increased in HEU at birth, but HEU NK cells have reduced IFNγ and perforin production, suggesting an adequate initial response, but decreased functional reserve. NK cell function improved with addition of exogenous IL-12, implicating impaired production of IL-12 by accessory cells. Alterations in NK cell and accessory

  17. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    International Nuclear Information System (INIS)

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

  18. Core-to-core uniformity improvement in multi-core fiber Bragg gratings

    Science.gov (United States)

    Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

    2014-07-01

    Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

  19. Analysis of the TREAT LEU Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-01

    Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

  20. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr