Sample records for power reactor 1400apr

  1. Human factors engineering verification and validation for APR1400 computerized control room

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

    Shin, Y. C.; Moon, H. K.; Kim, J. H.

    2006-07-01

    This paper introduces the Advanced Power Reactor 1400 (APR1400) HFE V and V activities the Korea Hydro Nuclear Plant Co. LTD. (KHNP) has performed for the last 10 years and some of the lessons learned through these activities. The features of APR1400 main control room include large display panel, redundant compact workstations, computer-based procedure, and safety console. Several iterations of human factors evaluations have been performed from small scale proof of concept tests to large scale integrated system tests for identifying human engineering deficiencies in the human system interface design. Evaluations in the proof of concept test were focused onmore » checking the presence of any show stopper problems in the design concept. Later evaluations were mostly for finding design problems and for assuring the resolution of human factors issues of advanced control room. The results of design evaluations were useful not only for refining the control room design, but also for licensing the standard design. Several versions of APR1400 mock-ups with dynamic simulation models of currently operating Korea Standard Nuclear Plant (KSNP) have been used for the evaluations with the participation of operators from KSNP plants. (authors)« less

  2. Reliability enhancement of APR + diverse protection system regarding common cause failures

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

    Oh, Y. G.; Kim, Y. M.; Yim, H. S.

    2012-07-01

    The Advanced Power Reactor Plus (APR +) nuclear power plant design has been developed on the basis of the APR1400 (Advanced Power Reactor 1400 MWe) to further enhance safety and economics. For the mitigation of Anticipated Transients Without Scram (ATWS) as well as Common Cause Failures (CCF) within the Plant Protection System (PPS) and the Emergency Safety Feature - Component Control System (ESF-CCS), several design improvement features have been implemented for the Diverse Protection System (DPS) of the APR + plant. As compared to the APR1400 DPS design, the APR + DPS has been designed to provide the Safety Injectionmore » Actuation Signal (SIAS) considering a large break LOCA accident concurrent with the CCF. Additionally several design improvement features, such as channel structure with redundant processing modules, and changes of system communication methods and auto-system test methods, are introduced to enhance the functional reliability of the DPS. Therefore, it is expected that the APR + DPS can provide an enhanced safety and reliability regarding possible CCF in the safety-grade I and C systems as well as the DPS itself. (authors)« less

  3. Use case driven approach to develop simulation model for PCS of APR1400 simulator

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

    Dong Wook, Kim; Hong Soo, Kim; Hyeon Tae, Kang

    2006-07-01

    The full-scope simulator is being developed to evaluate specific design feature and to support the iterative design and validation in the Man-Machine Interface System (MMIS) design of Advanced Power Reactor (APR) 1400. The simulator consists of process model, control logic model, and MMI for the APR1400 as well as the Power Control System (PCS). In this paper, a use case driven approach is proposed to develop a simulation model for PCS. In this approach, a system is considered from the point of view of its users. User's view of the system is based on interactions with the system and themore » resultant responses. In use case driven approach, we initially consider the system as a black box and look at its interactions with the users. From these interactions, use cases of the system are identified. Then the system is modeled using these use cases as functions. Lower levels expand the functionalities of each of these use cases. Hence, starting from the topmost level view of the system, we proceeded down to the lowest level (the internal view of the system). The model of the system thus developed is use case driven. This paper will introduce the functionality of the PCS simulation model, including a requirement analysis based on use case and the validation result of development of PCS model. The PCS simulation model using use case will be first used during the full-scope simulator development for nuclear power plant and will be supplied to Shin-Kori 3 and 4 plant. The use case based simulation model development can be useful for the design and implementation of simulation models. (authors)« less

  4. Response Time Analysis and Test of Protection System Instrument Channels for APR1400 and OPR1000

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

    Lee, Chang Jae; Han, Seung; Yun, Jae Hee

    2015-07-01

    Safety limits are required to maintain the integrity of physical barriers designed to prevent the uncontrolled release of radioactive materials in nuclear power plants. The safety analysis establishes two critical constraints that include an analytical limit in terms of a measured or calculated variable, and a specific time after the analytical limit is reached to begin protective action. Keeping with the nuclear regulations and industry standards, satisfying these two requirements will ensure that the safety limit will not be exceeded during the design basis event, either an anticipated operational occurrence or a postulated accident. Various studies on the setpoint determinationmore » methodology for the safety-related instrumentation have been actively performed to ensure that the requirement of the analytical limit is satisfied. In particular, the protection setpoint methodology for the advanced power reactor 1400 (APP1400) and the optimized power reactor 1000 (OPR1000) has been recently developed to cover both the design basis event and the beyond design basis event. The developed setpoint methodology has also been quantitatively validated using specific computer programs and setpoint calculations. However, the safety of nuclear power plants cannot be fully guaranteed by satisfying the requirement of the analytical limit. In spite of the response time verification requirements of nuclear regulations and industry standards, it is hard to find the studies on the systematically integrated methodology regarding the response time evaluation. In cases of APR1400 and OPR1000, the response time analysis for the plant protection system is partially included in the setpoint calculation and the response time test is separately performed via the specific plant procedure. The test technique has a drawback which is the difficulty to demonstrate completeness of timing test. The analysis technique has also a demerit of resulting in extreme times that not actually possible

  5. Main steam line break accident simulation of APR1400 using the model of ATLAS facility

    NASA Astrophysics Data System (ADS)

    Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

    2018-02-01

    A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

  6. Experimental and code simulation of a station blackout scenario for APR1400 with test facility ATLAS and MARS code

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

    Yu, X. G.; Kim, Y. S.; Choi, K. Y.

    2012-07-01

    A SBO (station blackout) experiment named SBO-01 was performed at full-pressure IET (Integral Effect Test) facility ATLAS (Advanced Test Loop for Accident Simulation) which is scaled down from the APR1400 (Advanced Power Reactor 1400 MWe). In this study, the transient of SBO-01 is discussed and is subdivided into three phases: the SG fluid loss phase, the RCS fluid loss phase, and the core coolant depletion and core heatup phase. In addition, the typical phenomena in SBO-01 test - SG dryout, natural circulation, core coolant boiling, the PRZ full, core heat-up - are identified. Furthermore, the SBO-01 test is reproduced bymore » the MARS code calculation with the ATLAS model which represents the ATLAS test facility. The experimental and calculated transients are then compared and discussed. The comparison reveals there was malfunction of equipments: the SG leakage through SG MSSV and the measurement error of loop flow meter. As the ATLAS model is validated against the experimental results, it can be further employed to investigate the other possible SBO scenarios and to study the scaling distortions in the ATLAS. (authors)« less

  7. Fuel assembly design for APR1400 with low CBC

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

    Hah, Chang Joo, E-mail: changhah@kings.ac.kr

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gdmore » rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.« less

  8. Advanced Design Features of APR1400 and Realization in Shin Kori Construction Project

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

    OH, S.J.; Park, K.C.; Kim, H.G.

    2006-07-01

    APR1400 adopted several advanced design features. To ensure their proper operation as a part of ShinKori 3,4 project, both experimental and analytical work are continuing. In this paper, work on the advanced design features related to enhanced safety is examined. APR1400 safety injection system consists of four independent trains which include four safety injection pump and tanks. A passive flow regulating device called fluidic device is installed in the safety injection tanks. Separate effect tests including a full scale fluidic device tests have been conducted. Integral system tests are in progress. Combination of these work with the analytical work usingmore » RELAP5/Mod3 would ensure the proper operation of the new safety injection systems. To mitigate severe accidents, hydrogen mitigation system using PARs and igniters is adopted. Also, active injection system and the streamlined insulation design are adopted to enhance the in-vessel retention capability with the external cooling of RPV strategy. Analytic work with supporting experiments is performed. We are certain that these preparatory work would help the successful adaptation of ADF in ShinKori project. (authors)« less

  9. A small, 1400 K, reactor for Brayton space power systems.

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

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

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

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

  11. 77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-03

    ... LLC reactor coolant equipment for four constructing four plant May 14, 2012 pumps with motors, APR1400... Emirates. XR176 monitoring and plant in Braka. 110060011 control equipment, auxiliary equipment and... NUCLEAR REGULATORY COMMISSION Application for a License To Export Nuclear Reactor Major Components...

  12. 14 CFR Appendix H to Part 23 - Installation of An Automatic Power Reserve (APR) System

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Installation of An Automatic Power Reserve (APR) System H Appendix H to Part 23 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT... AIRPLANES Pt. 23, App. H Appendix H to Part 23—Installation of An Automatic Power Reserve (APR) System H23.1...

  13. Defining the "proven technology" technical criterion in the reactor technology assessment for Malaysia's nuclear power program

    NASA Astrophysics Data System (ADS)

    Anuar, Nuraslinda; Kahar, Wan Shakirah Wan Abdul; Manan, Jamal Abdul Nasir Abd

    2015-04-01

    Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that "proven technology" is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for "proven technology" is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the "proven technology" term according to a specific country's requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of "proven technology" that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia's definition of "proven technology".

  14. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  15. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

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

  16. Thermionic reactors for space nuclear power

    NASA Technical Reports Server (NTRS)

    Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

    1985-01-01

    Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

  17. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

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

  18. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  19. Small and medium power reactors 1987

    NASA Astrophysics Data System (ADS)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

  20. Review of APR+ Level 2 PSA. Revision 2

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

    Lehner, John R.; Mubayi, Vinod; Pratt, W. Trevor

    2012-02-17

    Brookhaven National Laboratory (BNL) assisted the Korea Institute of Nuclear Safety (KINS) in reviewing the Level 2 Probabilistic Safety Assessment (PSA) of the APR+ Advanced Pressurized Water Reactor (PWR) prepared by the Korea Hydro & Nuclear Power Co., Ltd (KHNP) and KEPCO Engineering & Construction Co., Inc. (KEPCO-E&C). The work described in this report involves a review of the APR+ Level 2 PSA submittal [Ref. 1]. The PSA and, therefore, the review is limited to consideration of accidents initiated by internal events. As part of the review process, the review team also developed three sets of Requests for Additional Informationmore » (RAIs). These RAIs were provided to KHNP and KEPCO-E&C for their evaluation and response. This final detailed report documents the review findings for each technical element of the PSA and includes consideration of all of the RAIs made by the reviewers as well as the associated responses. This final report was preceded by an interim report [Ref. 2] that focused on identifying important issues regarding the PSA. In addition, a final meeting on the project was held at BNL on November 21-22, 2011, where BNL and KINS reviewers discussed their preliminary review findings with KHNP and KEPCO-E&C staffs. Additional information obtained during this final meeting was also used to inform the review findings of this final report. The review focused not only on the robustness of the APR+ design to withstand severe accidents, but also on the capability and acceptability of the Level 2 PSA in terms of level of detail and completeness. The Korean nuclear regulatory authorities will decide whether the PSA is acceptable and the BNL review team is providing its comments for KINS consideration. Section 2.0 provides the basis for the BNL review. Section 3.0 presents the review of each technical element of the PSA. Conclusions and a summary are presented in Section 4.0. Section 5.0 contains the references.« less

  1. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

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

  2. Small reactor power system for space application

    NASA Technical Reports Server (NTRS)

    Shirbacheh, M.

    1987-01-01

    A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

  3. 78 FR 14155 - Special Conditions: Learjet Inc., Model LJ-200-1A10 Airplane; Use of Automatic Power Reserve (APR...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-05

    ... Automatic Power Reserve (APR), an Automatic Takeoff Thrust Control System (ATTCS), for Go-Around Performance... airplane will have novel or unusual design features associated with utilizing go-around performance credit...: Federal eRegulations Portal: Go to http://www.regulations.gov/ and follow the online instructions for...

  4. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

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

  5. 21 CFR 882.1400 - Electroencephalograph.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Electroencephalograph. 882.1400 Section 882.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES NEUROLOGICAL DEVICES Neurological Diagnostic Devices § 882.1400 Electroencephalograph. (a...

  6. 21 CFR 882.1400 - Electroencephalograph.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Electroencephalograph. 882.1400 Section 882.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES NEUROLOGICAL DEVICES Neurological Diagnostic Devices § 882.1400 Electroencephalograph. (a...

  7. 7 CFR 1400.2 - Administration.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Administration. 1400.2 Section 1400.2 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT CORPORATION, DEPARTMENT OF... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.2 Administration. (a) The regulations in...

  8. 7 CFR 1400.2 - Administration.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 7 Agriculture 10 2013-01-01 2013-01-01 false Administration. 1400.2 Section 1400.2 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT CORPORATION, DEPARTMENT OF... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.2 Administration. (a) The regulations in...

  9. 7 CFR 1400.2 - Administration.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 7 Agriculture 10 2014-01-01 2014-01-01 false Administration. 1400.2 Section 1400.2 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT CORPORATION, DEPARTMENT OF... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.2 Administration. (a) The regulations in...

  10. 7 CFR 1400.2 - Administration.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 7 Agriculture 10 2012-01-01 2012-01-01 false Administration. 1400.2 Section 1400.2 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT CORPORATION, DEPARTMENT OF... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.2 Administration. (a) The regulations in...

  11. 7 CFR 1400.2 - Administration.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Administration. 1400.2 Section 1400.2 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT CORPORATION, DEPARTMENT OF... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.2 Administration. (a) The regulations in...

  12. 40 CFR 158.1400 - Definitions.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 24 2014-07-01 2014-07-01 false Definitions. 158.1400 Section 158.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS DATA REQUIREMENTS FOR PESTICIDES Residue Chemistry § 158.1400 Definitions. The following terms are defined for the...

  13. 40 CFR 158.1400 - Definitions.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 25 2013-07-01 2013-07-01 false Definitions. 158.1400 Section 158.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS DATA REQUIREMENTS FOR PESTICIDES Residue Chemistry § 158.1400 Definitions. The following terms are defined for the...

  14. 40 CFR 158.1400 - Definitions.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 25 2012-07-01 2012-07-01 false Definitions. 158.1400 Section 158.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS DATA REQUIREMENTS FOR PESTICIDES Residue Chemistry § 158.1400 Definitions. The following terms are defined for the...

  15. 7 CFR 1400.4 - Indian Tribe.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Indian Tribe. 1400.4 Section 1400.4 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.4 Indian Tribe. Provisions of this part do not apply to Indian tribes as defined in § 1400.3. ...

  16. 7 CFR 1400.4 - Indian Tribe.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Indian Tribe. 1400.4 Section 1400.4 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.4 Indian Tribe. Provisions of this part do not apply to Indian tribes as defined in § 1400.3. ...

  17. 7 CFR 1400.4 - Indian Tribe.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 7 Agriculture 10 2012-01-01 2012-01-01 false Indian Tribe. 1400.4 Section 1400.4 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.4 Indian Tribe. Provisions of this part do not apply to Indian tribes as defined in § 1400.3. ...

  18. 7 CFR 1400.4 - Indian Tribe.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 7 Agriculture 10 2013-01-01 2013-01-01 false Indian Tribe. 1400.4 Section 1400.4 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.4 Indian Tribe. Provisions of this part do not apply to Indian tribes as defined in § 1400.3. ...

  19. 7 CFR 1400.4 - Indian Tribe.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 7 Agriculture 10 2014-01-01 2014-01-01 false Indian Tribe. 1400.4 Section 1400.4 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.4 Indian Tribe. Provisions of this part do not apply to Indian tribes as defined in § 1400.3. ...

  20. 15 CFR 1400.6 - Construction.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 15 Commerce and Foreign Trade 3 2014-01-01 2014-01-01 false Construction. 1400.6 Section 1400.6 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.6 Construction. Nothing in...

  1. 15 CFR 1400.6 - Construction.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 15 Commerce and Foreign Trade 3 2012-01-01 2012-01-01 false Construction. 1400.6 Section 1400.6 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.6 Construction. Nothing in...

  2. 15 CFR 1400.6 - Construction.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 15 Commerce and Foreign Trade 3 2013-01-01 2013-01-01 false Construction. 1400.6 Section 1400.6 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.6 Construction. Nothing in...

  3. 15 CFR 1400.6 - Construction.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Construction. 1400.6 Section 1400.6 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.6 Construction. Nothing in...

  4. 15 CFR 1400.6 - Construction.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 15 Commerce and Foreign Trade 3 2011-01-01 2011-01-01 false Construction. 1400.6 Section 1400.6 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.6 Construction. Nothing in...

  5. 7 CFR 1400.205 - Trusts.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Trusts. 1400.205 Section 1400.205 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.205 Trusts. A trust will be considered to be actively engaged in farming with respect to a farming operation if: (a) The trust independently...

  6. 7 CFR 1400.500 - Applicability.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Average Adjusted Gross Income Limitation § 1400.500 Applicability... program payments or benefits described in § 1400.1 if the average adjusted gross income of the person or... average adjusted gross nonfarm income as defined in § 1400.3 that exceeds $500,000 will not be eligible to...

  7. 7 CFR 1400.500 - Applicability.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Average Adjusted Gross Income Limitation § 1400.500 Applicability... program payments or benefits described in § 1400.1 if the average adjusted gross income of the person or... average adjusted gross nonfarm income as defined in § 1400.3 that exceeds $500,000 will not be eligible to...

  8. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  9. 77 FR 60039 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-02

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

  10. 77 FR 38742 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-29

    ...-0087] RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission. ACTION... reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking input from the public, licensees, certificate...

  11. Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

    NASA Technical Reports Server (NTRS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

  12. Transmutation of actinides in power reactors.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  13. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision...

  14. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOEpatents

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  15. 7 CFR 1400.103 - Charitable organizations.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Charitable organizations. 1400.103 Section 1400.103... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.103 Charitable organizations. (a) A charitable organization, including a club, society, fraternal organization, or religious...

  16. PUSH-PULL POWER REACTOR

    DOEpatents

    Froman, D.K.

    1959-02-24

    Power generating nuclear reactors of the homogeneous liquid fuel type are discussed. The apparatus utilizes two identical reactors interconnected by conduits through heat exchanging apparatus. Each reactor contains a critical geometry region and a vapor region separated from the critical region by a baffle. When the liquid in the first critical region becomes critical, the vapor pressure above the fuel is increased due to the rise in the temperature until it forces the liquid fuel out of the first critical region through the heat exchanger and into the second critical region, which is at a lower temperature and consequently a lower vapor pressure. The above reaction is repeated in the second critical region and the liquid fuel is forced back into the first critical region. In this manner criticality is achieved alternately in each critical region and power is extracted by the heat exchanger from the liquid fuel passing therethrough. The vapor region and the heat exchanger have a non-critical geometry and reactivity control is effected by conventional control rods in the critical regions.

  17. Design of megawatt power level heat pipe reactors

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

    Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

  18. 7 CFR 1400.208 - Family members.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Family members. 1400.208 Section 1400.208 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.208 Family members. (a) Notwithstanding... persons, a majority of whom are family members, an adult family member who makes a significant...

  19. 7 CFR 1400.208 - Family members.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Family members. 1400.208 Section 1400.208 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.208 Family members. (a) Notwithstanding... persons, a majority of whom are family members, an adult family member who makes a significant...

  20. 7 CFR 1400.100 - Revocable trust.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Revocable trust. 1400.100 Section 1400.100... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.100 Revocable trust. A revocable trust and the grantor of the trust will be considered to be the same person. ...

  1. 7 CFR 1400.8 - Equitable treatment.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Equitable treatment. 1400.8 Section 1400.8... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.8 Equitable treatment. (a... Administrator deems necessary to provide fair and equitable treatment to such person or legal entity. (b...

  2. Gas-core reactor power transient analysis

    NASA Technical Reports Server (NTRS)

    Kascak, A. F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

  3. 7 CFR 1400.101 - Minor children.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Minor children. 1400.101 Section 1400.101 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.101 Minor children. (a) Except as provided in paragraph (b) of this section, payments received by a child under 18 years of age as of June 1...

  4. 7 CFR 1400.101 - Minor children.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 7 Agriculture 10 2012-01-01 2012-01-01 false Minor children. 1400.101 Section 1400.101 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.101 Minor children. (a) Except as provided in paragraph (b) of this section, payments received by a child under 18 years of age as of June 1...

  5. 7 CFR 1400.101 - Minor children.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 7 Agriculture 10 2014-01-01 2014-01-01 false Minor children. 1400.101 Section 1400.101 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.101 Minor children. (a) Except as provided in paragraph (b) of this section, payments received by a child under 18 years of age as of June 1...

  6. 7 CFR 1400.101 - Minor children.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Minor children. 1400.101 Section 1400.101 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.101 Minor children. (a) Except as provided in paragraph (b) of this section, payments received by a child under 18 years of age as of April 1...

  7. 7 CFR 1400.101 - Minor children.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 7 Agriculture 10 2013-01-01 2013-01-01 false Minor children. 1400.101 Section 1400.101 Agriculture... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.101 Minor children. (a) Except as provided in paragraph (b) of this section, payments received by a child under 18 years of age as of June 1...

  8. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

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

  9. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  10. 47 CFR 95.1400 - Basis and purpose.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false Basis and purpose. 95.1400 Section 95.1400 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES PERSONAL RADIO SERVICES Personal Locator Beacons (PLB). § 95.1400 Basis and purpose. The rules in this subpart are...

  11. 40 CFR 272.1400-272.1449 - [Reserved

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false [Reserved] 272.1400-272.1449 Section 272.1400-272.1449 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES (CONTINUED) APPROVED STATE HAZARDOUS WASTE MANAGEMENT PROGRAMS Nebraska §§ 272.1400-272.1449 [Reserved] ...

  12. 21 CFR 184.1400 - Lecithin.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Lecithin. 184.1400 Section 184.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) DIRECT FOOD SUBSTANCES AFFIRMED AS GENERALLY RECOGNIZED AS SAFE Listing of Specific...

  13. 21 CFR 184.1400 - Lecithin.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 3 2011-04-01 2011-04-01 false Lecithin. 184.1400 Section 184.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) DIRECT FOOD SUBSTANCES AFFIRMED AS GENERALLY RECOGNIZED AS SAFE Listing of Specific...

  14. 15 CFR 1400.1 - Purpose and scope.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Purpose and scope. 1400.1 Section 1400.1 Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) MINORITY BUSINESS DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.1 Purpose and...

  15. 29 CFR 1400.735-60 - Disciplinary actions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 4 2010-07-01 2010-07-01 false Disciplinary actions. 1400.735-60 Section 1400.735-60 Labor..., RESPONSIBILITIES, AND DISCIPLINE Disciplinary Actions and Penalties § 1400.735-60 Disciplinary actions. The Service shall take prompt disciplinary action against an employee committing prohibited activity, or whose...

  16. Power monitoring in space nuclear reactors using silicon carbide radiation detectors

    NASA Technical Reports Server (NTRS)

    Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

    2005-01-01

    Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

  17. 21 CFR 582.1400 - Lecithin.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Lecithin. 582.1400 Section 582.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS... Lecithin. (a) Product. Lecithin. (b) Conditions of use. This substance is generally recognized as safe when...

  18. Thermionic reactor power conditioner design for nuclear electric propulsion.

    NASA Technical Reports Server (NTRS)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  19. Deployment history and design considerations for space reactor power systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.

    2009-05-01

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

  20. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    NASA Astrophysics Data System (ADS)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  1. Reference Reactor Module for the Affordable Fission Surface Power System

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  2. Consumption of the electric power inside silent discharge reactors

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

    Yehia, Ashraf, E-mail: yehia30161@yahoo.com

    An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodesmore » in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.« less

  3. Reactor engineering support of operations at the Davis-Besse nuclear power station

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

    Kelley, D.B.

    1995-12-31

    Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

  4. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

  5. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

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

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  6. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

  7. Long lifetime fast spectrum reactor for lunar surface power system

    NASA Astrophysics Data System (ADS)

    Kambe, Mitsuru

    1993-01-01

    In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

  8. 40 CFR 1400.12 - Qualified researchers.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Qualified researchers. 1400.12 Section 1400.12 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7...

  9. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  10. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  11. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  12. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  13. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  14. COST FUNCTION STUDIES FOR POWER REACTORS

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

    Heestand, J.; Wos, L.T.

    1961-11-01

    A function to evaluate the cost of electricity produced by a nuclear power reactor was developed. The basic equation, revenue = capital charges + profit + operating expenses, was expanded in terms of various cost parameters to enable analysis of multiregion nuclear reactors with uranium and/or plutonium for fuel. A corresponding IBM 704 computer program, which will compute either the price of electricity or the value of plutonium, is presented in detail. (auth)

  15. Reference reactor module for NASA's lunar surface fission power system

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

    Poston, David I; Kapernick, Richard J; Dixon, David D

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on themore » lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.« less

  16. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-01

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  17. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

  18. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

  19. 40 CFR 1400.7 - In general.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false In general. 1400.7 Section 1400.7 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF...

  20. Design and simulation of a novel 1400 V-4000 V enhancement mode buried gate GaN HEMT for power applications

    NASA Astrophysics Data System (ADS)

    Faramehr, Soroush; Kalna, Karol; Igić, Petar

    2014-11-01

    A novel enhancement mode structure, a buried gate gallium nitride (GaN) high electron mobility transistor (HEMT) with a breakdown voltage (BV) of 1400 V-4000 V for a source-to-drain spacing (LSD) of 6 μm-32 μm, is investigated using simulations by Silvaco Atlas. The simulations are based on meticulous calibration of a conventional lateral 1 μm gate length GaN HEMT with a source-to-drain spacing of 6 μm against its experimental transfer characteristics and BV. The specific on-resistance RS for the new power transistor with the source-to-drain spacing of 6 μm showing BV = 1400 V and the source-to-drain spacing of 8 μm showing BV = 1800 V is found to be 2.3 mΩ · cm2 and 3.5 mΩ · cm2, respectively. Further improvement up to BV = 4000 V can be achieved by increasing the source-to-drain spacing to 32 μm with the specific on-resistance of RS = 35.5 mΩ · cm2. The leakage current in the proposed devices stays in the range of ˜5 × 10-9 mA mm-1.

  1. 46 CFR 154.1400 - Safety equipment: All vessels.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Safety equipment: All vessels. 154.1400 Section 154.1400 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR SELF-PROPELLED VESSELS CARRYING BULK LIQUEFIED GASES Design, Construction and Equipment Safety Equipment § 154.1400 Safety...

  2. 46 CFR 154.1400 - Safety equipment: All vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 5 2011-10-01 2011-10-01 false Safety equipment: All vessels. 154.1400 Section 154.1400 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR SELF-PROPELLED VESSELS CARRYING BULK LIQUEFIED GASES Design, Construction and Equipment Safety Equipment § 154.1400 Safety...

  3. 21 CFR 868.1400 - Carbon dioxide gas analyzer.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Carbon dioxide gas analyzer. 868.1400 Section 868.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Diagnostic Devices § 868.1400 Carbon dioxide gas analyzer. (a) Identification. A carbon dioxide gas analyzer...

  4. 7 CFR 1400.212 - Growers of hybrid seed.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Growers of hybrid seed. 1400.212 Section 1400.212... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.212 Growers of hybrid seed. The existence of a hybrid seed contract for a person or legal entity will not be taken into account...

  5. 7 CFR 1400.212 - Growers of hybrid seed.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Growers of hybrid seed. 1400.212 Section 1400.212... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.212 Growers of hybrid seed. The existence of a hybrid seed contract for a person or legal entity will not be taken into account...

  6. 21 CFR 876.1400 - Stomach pH electrode.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Stomach pH electrode. 876.1400 Section 876.1400...) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Diagnostic Devices § 876.1400 Stomach pH electrode. (a) Identification. A stomach pH electrode is a device used to measure intragastric and intraesophageal pH (hydrogen...

  7. 21 CFR 876.1400 - Stomach pH electrode.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Stomach pH electrode. 876.1400 Section 876.1400...) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Diagnostic Devices § 876.1400 Stomach pH electrode. (a) Identification. A stomach pH electrode is a device used to measure intragastric and intraesophageal pH (hydrogen...

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

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

    John Darrell Bess

    2008-06-01

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

  9. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  10. Gas core reactors for actinide transmutation. [uranium hexafluoride

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.; Wan, P. T.; Chow, S.

    1979-01-01

    The preliminary design of a uranium hexafluoride actinide transmutation reactor to convert long-lived actinide wastes to shorter-lived fission product wastes was analyzed. It is shown that externally moderated gas core reactors are ideal radiators. They provide an abundant supply of thermal neutrons and are insensitive to composition changes in the blanket. For the present reactor, an initial load of 6 metric tons of actinides is loaded. This is equivalent to the quantity produced by 300 LWR-years of operation. At the beginning, the core produces 2000 MWt while the blanket generates only 239 MWt. After four years of irradiation, the actinide mass is reduced to 3.9 metric tonnes. During this time, the blanket is becoming more fissile and its power rapidly approaches 1600 MWt. At the end of four years, continuous refueling of actinides is carried out and the actinide mass is held constant. Equilibrium is essentially achieved at the end of eight years. At equilibrium, the core is producing 1400 MWt and the blanket 1600 MWt. At this power level, the actinide destruction rate is equal to the production rate from 32 LWRs.

  11. 21 CFR 876.1400 - Stomach pH electrode.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... ion concentration). The pH electrode is at the end of a flexible lead which may be inserted into the... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Stomach pH electrode. 876.1400 Section 876.1400...) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Diagnostic Devices § 876.1400 Stomach pH electrode. (a...

  12. 21 CFR 876.1400 - Stomach pH electrode.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... ion concentration). The pH electrode is at the end of a flexible lead which may be inserted into the... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Stomach pH electrode. 876.1400 Section 876.1400...) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Diagnostic Devices § 876.1400 Stomach pH electrode. (a...

  13. 21 CFR 876.1400 - Stomach pH electrode.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... ion concentration). The pH electrode is at the end of a flexible lead which may be inserted into the... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Stomach pH electrode. 876.1400 Section 876.1400...) MEDICAL DEVICES GASTROENTEROLOGY-UROLOGY DEVICES Diagnostic Devices § 876.1400 Stomach pH electrode. (a...

  14. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-15

    ... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

  15. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  16. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-09

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

  17. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    DOEpatents

    Hampel, Viktor E.

    1989-01-01

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

  18. Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

    Jung, Y. S.; Joo, H. G.; Yoon, J. I.

    The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

  19. Nuclear design of a vapor core reactor for space nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

    1993-01-01

    Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

  20. 48 CFR 22.1400 - Scope of subpart.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Scope of subpart. 22.1400 Section 22.1400 Federal Acquisition Regulations System FEDERAL ACQUISITION REGULATION SOCIOECONOMIC PROGRAMS APPLICATION OF LABOR LAWS TO GOVERNMENT ACQUISITIONS Employment of Workers with Disabilities 22...

  1. 12 CFR 1400.1 - Farm Credit System Insurance Corporation.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 12 Banks and Banking 9 2013-01-01 2013-01-01 false Farm Credit System Insurance Corporation. 1400.1 Section 1400.1 Banks and Banking FARM CREDIT SYSTEM INSURANCE CORPORATION ORGANIZATION AND FUNCTIONS Organization and Functions § 1400.1 Farm Credit System Insurance Corporation. The Farm Credit...

  2. 12 CFR 1400.1 - Farm Credit System Insurance Corporation.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 12 Banks and Banking 9 2012-01-01 2012-01-01 false Farm Credit System Insurance Corporation. 1400.1 Section 1400.1 Banks and Banking FARM CREDIT SYSTEM INSURANCE CORPORATION ORGANIZATION AND FUNCTIONS Organization and Functions § 1400.1 Farm Credit System Insurance Corporation. The Farm Credit...

  3. 12 CFR 1400.1 - Farm Credit System Insurance Corporation.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 12 Banks and Banking 7 2011-01-01 2011-01-01 false Farm Credit System Insurance Corporation. 1400.1 Section 1400.1 Banks and Banking FARM CREDIT SYSTEM INSURANCE CORPORATION ORGANIZATION AND FUNCTIONS Organization and Functions § 1400.1 Farm Credit System Insurance Corporation. The Farm Credit...

  4. 12 CFR 1400.1 - Farm Credit System Insurance Corporation.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 12 Banks and Banking 10 2014-01-01 2014-01-01 false Farm Credit System Insurance Corporation. 1400.1 Section 1400.1 Banks and Banking FARM CREDIT SYSTEM INSURANCE CORPORATION ORGANIZATION AND FUNCTIONS Organization and Functions § 1400.1 Farm Credit System Insurance Corporation. The Farm Credit...

  5. 12 CFR 1400.1 - Farm Credit System Insurance Corporation.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 12 Banks and Banking 7 2010-01-01 2010-01-01 false Farm Credit System Insurance Corporation. 1400.1 Section 1400.1 Banks and Banking FARM CREDIT SYSTEM INSURANCE CORPORATION ORGANIZATION AND FUNCTIONS Organization and Functions § 1400.1 Farm Credit System Insurance Corporation. The Farm Credit...

  6. 29 CFR 1400.735-3 - Advice and counseling service.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 29 Labor 4 2012-07-01 2012-07-01 false Advice and counseling service. 1400.735-3 Section 1400.735-3 Labor Regulations Relating to Labor (Continued) FEDERAL MEDIATION AND CONCILIATION SERVICE STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE General § 1400.735-3 Advice and counseling service...

  7. 29 CFR 1400.735-3 - Advice and counseling service.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 4 2010-07-01 2010-07-01 false Advice and counseling service. 1400.735-3 Section 1400.735-3 Labor Regulations Relating to Labor (Continued) FEDERAL MEDIATION AND CONCILIATION SERVICE STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE General § 1400.735-3 Advice and counseling service...

  8. 29 CFR 1400.735-3 - Advice and counseling service.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 29 Labor 4 2013-07-01 2013-07-01 false Advice and counseling service. 1400.735-3 Section 1400.735-3 Labor Regulations Relating to Labor (Continued) FEDERAL MEDIATION AND CONCILIATION SERVICE STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE General § 1400.735-3 Advice and counseling service...

  9. 29 CFR 1400.735-3 - Advice and counseling service.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 29 Labor 4 2014-07-01 2014-07-01 false Advice and counseling service. 1400.735-3 Section 1400.735-3 Labor Regulations Relating to Labor (Continued) FEDERAL MEDIATION AND CONCILIATION SERVICE STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE General § 1400.735-3 Advice and counseling service...

  10. 29 CFR 1400.735-3 - Advice and counseling service.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 29 Labor 4 2011-07-01 2011-07-01 false Advice and counseling service. 1400.735-3 Section 1400.735-3 Labor Regulations Relating to Labor (Continued) FEDERAL MEDIATION AND CONCILIATION SERVICE STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE General § 1400.735-3 Advice and counseling service...

  11. Fuel supply of nuclear power industry with the introduction of fast reactors

    NASA Astrophysics Data System (ADS)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  12. Non-equilibrium plasma reactors for organic solvent destruction

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

    Yang, C.L.; Beltran, M.R.; Kravets, Z.

    1997-12-31

    Two non-equilibrium plasma reactors were evaluated for their ability to destroy three widely used organic solvents, i.e., 2-butanone, toluene and ethyl acetate. The catalyzed plasma reactor (CPR) with 6 mm glass beads destroys 98% of 50 ppm toluene in air at 24 kV/cm and space velocities of 1,400 v/v/hr. Eighty-five percent of ethyl acetate and 2-butanone are destroyed under the same conditions. The tubular plasma reactor (TPR) has an efficiency of 10% to 20% lower than that of a CPR under the same conditions. The 1,400 v/v/hr in a CPR is equal to a residence time of 2.6 seconds inmore » a TPR. The operating temperatures, corona characteristics, as well as the kinetics of VOC destruction in both TPR and CPR were studied.« less

  13. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    PubMed

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-10-09

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  14. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    NASA Technical Reports Server (NTRS)

    George, Jeffrey A.

    1991-01-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  15. Closed Brayton cycle power conversion systems for nuclear reactors :

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

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors,more » reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  16. 40 CFR 1400.13 - Read-only database.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Read-only database. 1400.13 Section 1400.13 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7...

  17. 40 CFR 1400.6 - Enhanced local access.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Enhanced local access. 1400.6 Section 1400.6 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7...

  18. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    PubMed Central

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  19. 21 CFR 862.1400 - Hydroxyproline test system.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... to measure the amino acid hydroxyproline in urine. Hydroxyproline measurements are used in the... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Hydroxyproline test system. 862.1400 Section 862.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED...

  20. 21 CFR 862.1400 - Hydroxyproline test system.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... to measure the amino acid hydroxyproline in urine. Hydroxyproline measurements are used in the... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Hydroxyproline test system. 862.1400 Section 862.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED...

  1. 21 CFR 862.1400 - Hydroxyproline test system.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Hydroxyproline test system. 862.1400 Section 862.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES CLINICAL CHEMISTRY AND CLINICAL TOXICOLOGY DEVICES Clinical Chemistry Test Systems § 862...

  2. 40 CFR 1400.13 - Read-only database.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 34 2012-07-01 2012-07-01 false Read-only database. 1400.13 Section... INFORMATION Other Provisions § 1400.13 Read-only database. The Administrator is authorized to establish... public off-site consequence analysis information by means of a central database under the control of the...

  3. 40 CFR 1400.13 - Read-only database.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 33 2014-07-01 2014-07-01 false Read-only database. 1400.13 Section... INFORMATION Other Provisions § 1400.13 Read-only database. The Administrator is authorized to establish... public off-site consequence analysis information by means of a central database under the control of the...

  4. 40 CFR 1400.13 - Read-only database.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 33 2011-07-01 2011-07-01 false Read-only database. 1400.13 Section... INFORMATION Other Provisions § 1400.13 Read-only database. The Administrator is authorized to establish... public off-site consequence analysis information by means of a central database under the control of the...

  5. 40 CFR 1400.13 - Read-only database.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 34 2013-07-01 2013-07-01 false Read-only database. 1400.13 Section... INFORMATION Other Provisions § 1400.13 Read-only database. The Administrator is authorized to establish... public off-site consequence analysis information by means of a central database under the control of the...

  6. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

  7. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

  8. Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

    NASA Astrophysics Data System (ADS)

    Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

  9. Performance Assessment of the Commercial CFD Software for the Prediction of the Reactor Internal Flow

    NASA Astrophysics Data System (ADS)

    Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong; Kim, Do Hyeong; Kang, Min Ku

    2014-06-01

    As the computer hardware technology develops the license applicants for nuclear power plant use the commercial CFD software with the aim of reducing the excessive conservatism associated with using simplified and conservative analysis tools. Even if some of CFD software developer and its user think that a state of the art CFD software can be used to solve reasonably at least the single-phase nuclear reactor problems, there is still limitation and uncertainty in the calculation result. From a regulatory perspective, Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of the commercial CFD software for nuclear reactor problems. In this study, in order to examine the validity of the results of 1/5 scaled APR+ (Advanced Power Reactor Plus) flow distribution tests and the applicability of CFD in the analysis of reactor internal flow, the simulation was conducted with the two commercial CFD software (ANSYS CFX V.14 and FLUENT V.14) among the numerous commercial CFD software and was compared with the measurement. In addition, what needs to be improved in CFD for the accurate simulation of reactor core inlet flow was discussed.

  10. 45 CFR 155.1400 - Quality rating system.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... the quality rating information assigned to each QHP on its Web site, in accordance with § 155.205(b)(1... 45 Public Welfare 1 2014-10-01 2014-10-01 false Quality rating system. 155.1400 Section 155.1400... EXCHANGE ESTABLISHMENT STANDARDS AND OTHER RELATED STANDARDS UNDER THE AFFORDABLE CARE ACT Quality...

  11. A facility for testing 10 to 100-kWe space power reactors

    NASA Astrophysics Data System (ADS)

    Carlson, William F.; Bitten, Ernest J.

    1993-01-01

    This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

  12. Adaptive control method for core power control in TRIGA Mark II reactor

    NASA Astrophysics Data System (ADS)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  13. An underground nuclear power station using self-regulating heat-pipe controlled reactors

    DOEpatents

    Hampel, V.E.

    1988-05-17

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

  14. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

  15. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

  16. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

  17. FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

    Lipinski, R.J.; Pickard, P.S.

    1992-12-31

    FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

  18. FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

    Lipinski, R.J.; Pickard, P.S.

    1992-01-01

    FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

  19. 13 CFR 120.1400 - Grounds for enforcement actions-SBA Lenders.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Grounds for enforcement actions-SBA Lenders. 120.1400 Section 120.1400 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Risk-Based Lender Oversight Enforcement Actions § 120.1400 Grounds for enforcement actions...

  20. ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

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

    Mahlmeister, J E; Haberer, W V; Casey, D F

    1960-12-15

    The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

  1. Summary of space nuclear reactor power systems, 1983 - 1992

    NASA Astrophysics Data System (ADS)

    Buden, D.

    1993-08-01

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  2. 40 CFR 1400.10 - Limitation on public dissemination.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Limitation on public dissemination. 1400.10 Section 1400.10 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT...

  3. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

  4. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

  5. Gaseous-fuel nuclear reactor research for multimegawatt power in space

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T.; Helmick, H. H.

    1977-01-01

    In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

  6. 21 CFR 892.1400 - Nuclear sealed calibration source.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... reference radionuclide intended for calibration of medical nuclear radiation detectors. (b) Classification... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Nuclear sealed calibration source. 892.1400... (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1400 Nuclear sealed calibration source...

  7. 21 CFR 892.1400 - Nuclear sealed calibration source.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... reference radionuclide intended for calibration of medical nuclear radiation detectors. (b) Classification... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Nuclear sealed calibration source. 892.1400... (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1400 Nuclear sealed calibration source...

  8. 21 CFR 892.1400 - Nuclear sealed calibration source.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... reference radionuclide intended for calibration of medical nuclear radiation detectors. (b) Classification... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Nuclear sealed calibration source. 892.1400... (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1400 Nuclear sealed calibration source...

  9. 21 CFR 892.1400 - Nuclear sealed calibration source.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... reference radionuclide intended for calibration of medical nuclear radiation detectors. (b) Classification... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Nuclear sealed calibration source. 892.1400... (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1400 Nuclear sealed calibration source...

  10. 21 CFR 892.1400 - Nuclear sealed calibration source.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... reference radionuclide intended for calibration of medical nuclear radiation detectors. (b) Classification... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Nuclear sealed calibration source. 892.1400... (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1400 Nuclear sealed calibration source...

  11. NACA Zero Power Reactor Facility Hazards Summary

    NASA Technical Reports Server (NTRS)

    1957-01-01

    The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

  12. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

  13. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

  14. 7 CFR 1400.7 - Commensurate contributions and risk.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false Commensurate contributions and risk. 1400.7 Section 1400.7 Agriculture Regulations of the Department of Agriculture (Continued) COMMODITY CREDIT... contributions and risk. (a) In order to be considered eligible to receive payments under the programs specified...

  15. 40 CFR 1400.4 - Vulnerable zone indicator system.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Vulnerable zone indicator system. 1400.4 Section 1400.4 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT PROGRAMS UNDER THE CLEAN AIR ACT SECTION 112(r)(7...

  16. 10 CFR 2.1400 - Purpose and scope of subpart N.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Purpose and scope of subpart N. 2.1400 Section 2.1400... ORDERS Expedited Proceedings with Oral Hearings § 2.1400 Purpose and scope of subpart N. The purpose of... parties otherwise agree and request the application of subpart N procedures, and proceedings for the...

  17. 10 CFR 2.1400 - Purpose and scope of subpart N.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Purpose and scope of subpart N. 2.1400 Section 2.1400... ORDERS Expedited Proceedings with Oral Hearings § 2.1400 Purpose and scope of subpart N. The purpose of... parties otherwise agree and request the application of subpart N procedures, and proceedings for the...

  18. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    NASA Astrophysics Data System (ADS)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

  19. Autonomous Control Capabilities for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  20. 76 FR 74630 - Making Changes to Emergency Plans for Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-01

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 52 RIN 3150-AI10 [NRC-2008-0122] Making Changes to Emergency Plans for Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... guide (RG) 1.219, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors.'' This...

  1. 7 CFR 1400.209 - Sharecroppers.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.209 Sharecroppers. (a) Notwithstanding...

  2. Summary of space nuclear reactor power systems, 1983--1992

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

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

  3. 21 CFR 868.1400 - Carbon dioxide gas analyzer.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Carbon dioxide gas analyzer. 868.1400 Section 868...) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Diagnostic Devices § 868.1400 Carbon dioxide gas analyzer. (a) Identification. A carbon dioxide gas analyzer is a device intended to measure the concentration of carbon dioxide...

  4. 21 CFR 868.1400 - Carbon dioxide gas analyzer.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Carbon dioxide gas analyzer. 868.1400 Section 868...) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Diagnostic Devices § 868.1400 Carbon dioxide gas analyzer. (a) Identification. A carbon dioxide gas analyzer is a device intended to measure the concentration of carbon dioxide...

  5. 21 CFR 868.1400 - Carbon dioxide gas analyzer.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Carbon dioxide gas analyzer. 868.1400 Section 868...) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Diagnostic Devices § 868.1400 Carbon dioxide gas analyzer. (a) Identification. A carbon dioxide gas analyzer is a device intended to measure the concentration of carbon dioxide...

  6. 21 CFR 868.1400 - Carbon dioxide gas analyzer.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 8 2014-04-01 2014-04-01 false Carbon dioxide gas analyzer. 868.1400 Section 868...) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Diagnostic Devices § 868.1400 Carbon dioxide gas analyzer. (a) Identification. A carbon dioxide gas analyzer is a device intended to measure the concentration of carbon dioxide...

  7. 40 CFR 1400.4 - Vulnerable zone indicator system.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 40 Protection of Environment 33 2014-07-01 2014-07-01 false Vulnerable zone indicator system. 1400... INFORMATION Public Access § 1400.4 Vulnerable zone indicator system. (a) In general. The Administrator shall provide access to a computer-based indicator that shall inform any person located in any state whether an...

  8. 40 CFR 1400.4 - Vulnerable zone indicator system.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 34 2012-07-01 2012-07-01 false Vulnerable zone indicator system. 1400... INFORMATION Public Access § 1400.4 Vulnerable zone indicator system. (a) In general. The Administrator shall provide access to a computer-based indicator that shall inform any person located in any state whether an...

  9. 40 CFR 1400.4 - Vulnerable zone indicator system.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 34 2013-07-01 2013-07-01 false Vulnerable zone indicator system. 1400... INFORMATION Public Access § 1400.4 Vulnerable zone indicator system. (a) In general. The Administrator shall provide access to a computer-based indicator that shall inform any person located in any state whether an...

  10. 40 CFR 1400.4 - Vulnerable zone indicator system.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 33 2011-07-01 2011-07-01 false Vulnerable zone indicator system. 1400... INFORMATION Public Access § 1400.4 Vulnerable zone indicator system. (a) In general. The Administrator shall provide access to a computer-based indicator that shall inform any person located in any state whether an...

  11. 46 CFR 154.1400 - Safety equipment: All vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... § 154.1). (5) Three helmets that meet ANSI Safety Requirements for Industrial Head Protection, Z-89.1... 46 Shipping 5 2012-10-01 2012-10-01 false Safety equipment: All vessels. 154.1400 Section 154.1400 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY...

  12. 46 CFR 154.1400 - Safety equipment: All vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... § 154.1). (5) Three helmets that meet ANSI Safety Requirements for Industrial Head Protection, Z-89.1... 46 Shipping 5 2013-10-01 2013-10-01 false Safety equipment: All vessels. 154.1400 Section 154.1400 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY...

  13. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

  14. EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

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

    Alexander, L.G.; Kinyon, B.W.; Lackey, M.E.

    1960-03-24

    A preliminary design study was made of an experimental molten-salt- fueled power reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler steam-generating cycle. Conceptual plant layouts, basic information on the major fuel circuit components, a process flowsheet, and the nuclear characteristics of the core are presented. The design plant electrical output is 10 Mw, and the total construction cost is estimated to be approximately ,000,000. (auth)

  15. Background radiation measurements at high power research reactors

    NASA Astrophysics Data System (ADS)

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zhang, C.; Zhang, X.; Prospect Collaboration

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  16. 7 CFR 1400.202 - Persons.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.202 Persons. (a) A person will be...

  17. 7 CFR 1400.207 - Landowners.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.207 Landowners. (a) A person or legal...

  18. Low-power lead-cooled fast reactor loaded with MOX-fuel

    NASA Astrophysics Data System (ADS)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  19. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

    Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

    1973-01-01

    A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

  20. REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

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

    Appleby, E.R., comp

    1963-10-01

    Unclassified information from domestic and foreign literature from January 1952 through September 1963 is compiled. Design characteristics and current information on the status of the individual designs are given, along with references for the associated literature. SNAP systems, proposed reactors, and chemonuclear and test reactors with characteristics similar to power reactors are included. The designs are indexed by name, location, type, and some special characteristics. (D.C.W.)

  1. Nuclear reactor power for an electrically powered orbital transfer vehicle

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  2. 7 CFR 1400.301 - Eligibility.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Cash Rent Tenants § 1400.301 Eligibility. (a) Any tenant that is...

  3. 7 CFR 1400.9 - Appeals.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.9 Appeals. (a) A person or legal entity...

  4. Automated power control system for reactor TRIGA PUSPATI

    NASA Astrophysics Data System (ADS)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

    2017-01-01

    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  5. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  6. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  7. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  8. 7 CFR 1400.401 - Eligibility.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Foreign Persons § 1400.401 Eligibility. (a) Any person who is not a...

  9. 47 CFR 73.1400 - Transmission system monitoring and control.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ....1400 Section 73.1400 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) BROADCAST RADIO... unnecessary. (b) Unattended operation. Unattended operation is either the absence of human supervision or the substitution of automated supervision of a station's transmission system for human supervision. In the former...

  10. NASA's Kilopower Reactor Development and the Path to Higher Power Missions

    NASA Technical Reports Server (NTRS)

    Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

    2017-01-01

    The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

  11. NASA's Kilopower Reactor Development and the Path to Higher Power Missions

    NASA Technical Reports Server (NTRS)

    Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

    2017-01-01

    The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

  12. 30 CFR 75.1400-4 - Certifications and records of daily examinations.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... by § 75.1400-3, the person conducting the examination shall make a record of the condition and the... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Certifications and records of daily examinations. 75.1400-4 Section 75.1400-4 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT...

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

    NASA Technical Reports Server (NTRS)

    Davis, H. S.

    1984-01-01

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

  14. 21 CFR 184.1400 - Lecithin.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ....1400 Lecithin. (a) Commercial lecithin is a naturally occurring mixture of the phosphatides of choline, ethanolamine, and inositol, with smaller amounts of othe lipids. It is isolated as a gum following hydration of...

  15. Background radiation measurements at high power research reactors

    DOE PAGES

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; ...

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  16. 2 CFR 1400.20 - When does this part apply to me?

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 2 Grants and Agreements 1 2012-01-01 2012-01-01 false When does this part apply to me? 1400.20 Section 1400.20 Grants and Agreements Federal Agency Regulations for Grants and Agreements DEPARTMENT OF THE INTERIOR NONPROCUREMENT DEBARMENT AND SUSPENSION § 1400.20 When does this part apply to me? This...

  17. 2 CFR 1400.20 - When does this part apply to me?

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 2 Grants and Agreements 1 2013-01-01 2013-01-01 false When does this part apply to me? 1400.20 Section 1400.20 Grants and Agreements Federal Agency Regulations for Grants and Agreements DEPARTMENT OF THE INTERIOR NONPROCUREMENT DEBARMENT AND SUSPENSION § 1400.20 When does this part apply to me? This...

  18. 2 CFR 1400.20 - When does this part apply to me?

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 2 Grants and Agreements 1 2010-01-01 2010-01-01 false When does this part apply to me? 1400.20 Section 1400.20 Grants and Agreements Federal Agency Regulations for Grants and Agreements DEPARTMENT OF THE INTERIOR NONPROCUREMENT DEBARMENT AND SUSPENSION § 1400.20 When does this part apply to me? This...

  19. 2 CFR 1400.20 - When does this part apply to me?

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 2 Grants and Agreements 1 2011-01-01 2011-01-01 false When does this part apply to me? 1400.20 Section 1400.20 Grants and Agreements Federal Agency Regulations for Grants and Agreements DEPARTMENT OF THE INTERIOR NONPROCUREMENT DEBARMENT AND SUSPENSION § 1400.20 When does this part apply to me? This...

  20. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

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

    1976-01-01

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

  1. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  2. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-16

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

  3. 75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-20

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

  4. 21 CFR 177.1400 - Hydroxyethyl cellulose film, water-insoluble.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... packaging food in accordance with the following prescribed conditions: (a) Water-insoluble hydroxyethyl... 21 Food and Drugs 3 2013-04-01 2013-04-01 false Hydroxyethyl cellulose film, water-insoluble. 177.1400 Section 177.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN...

  5. 21 CFR 177.1400 - Hydroxyethyl cellulose film, water-insoluble.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... packaging food in accordance with the following prescribed conditions: (a) Water-insoluble hydroxyethyl... 21 Food and Drugs 3 2012-04-01 2012-04-01 false Hydroxyethyl cellulose film, water-insoluble. 177.1400 Section 177.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN...

  6. Evaluation of power density on the bioethanol production using mesoscale oscillatory baffled reactor and stirred tank reactor

    NASA Astrophysics Data System (ADS)

    Yussof, H. W.; Bahri, S. S.; Mazlan, N. A.

    2018-03-01

    A recent development in oscillatory baffled reactor technology is down-scaling the reactor, so that it can be used for production of small-scale bioproduct. In the present study, a mesoscale oscillatory baffled reactor (MOBR) with central baffle system was developed. The reactor performance of the MOBR was compared with conventional stirred tank reactor (STR) to evaluate the performance of bioethanol fermentation using Saccharomyces cerevisiae. Evaluation was made at similar power density of 24.21, 57.38, 112.35 and 193.67 Wm-3 by varying frequency (f), amplitude (xo) and agitation speed (rpm). It was found that the MOBR improved the mixing intensity resulted in lower glucose concentration (0.988 gL-1) and higher bioethanol concentration (38.98 gL-1) after 12 hours fermentation at power density of 193.67 Wm-3. Based on the results, the bioethanol yield obtained using MOBR was 39% higher than the maximum achieved in STR. Bioethanol production using MOBR proved to be feasible as it is not only able to compete with conventional STR but also offers advantages of straight-forward scale-up, whereas it is complicated and difficult in STR. Overall, MOBR offers great prospective over the conventional STR.

  7. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

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

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

  8. The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

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

    Hollaway, W.R.

    1991-08-01

    If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

  9. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    NASA Technical Reports Server (NTRS)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  10. Oxygen transport membrane reactor based method and system for generating electric power

    DOEpatents

    Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

    2017-02-07

    A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

  11. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  12. Preliminary plan for testing a thermionic reactor in the Plum Brook Space Power Facility

    NASA Technical Reports Server (NTRS)

    Haley, F. A.

    1972-01-01

    A preliminary plan is presented for testing a thermionic reactor in the Plum Brook Space Power Facility (SPF). A technical approach, cost estimate, manpower estimate, and schedule are presented to cover a 2 year full power reactor test.

  13. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

  14. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

  15. Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

    NASA Technical Reports Server (NTRS)

    Silverman, S. W.; Willenberg, H. J.; Robertson, C.

    1985-01-01

    An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

  16. 77 FR 40092 - License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company, St...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-06

    ... Increase the Maximum Reactor Power Level, Florida Power & Light Company, St. Lucie, Units 1 and 2 AGENCY... amendment for Renewed Facility Operating License Nos. DPR-67 and NPF-16, issued to Florida Power & Light... St. Lucie County, Florida. The proposed license amendment would increase the maximum thermal power...

  17. 20 CFR 416.1400 - Introduction.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... resolved. (b) Nature of the administrative review process. In making a determination or decision in your... DISABLED Determinations, Administrative Review Process, and Reopening of Determinations and Decisions Introduction, Definitions, and Initial Determinations § 416.1400 Introduction. (a) Explanation of the...

  18. Predicting High-Power Performance in Professional Cyclists.

    PubMed

    Sanders, Dajo; Heijboer, Mathieu; Akubat, Ibrahim; Meijer, Kenneth; Hesselink, Matthijs K

    2017-03-01

    To assess if short-duration (5 to ~300 s) high-power performance can accurately be predicted using the anaerobic power reserve (APR) model in professional cyclists. Data from 4 professional cyclists from a World Tour cycling team were used. Using the maximal aerobic power, sprint peak power output, and an exponential constant describing the decrement in power over time, a power-duration relationship was established for each participant. To test the predictive accuracy of the model, several all-out field trials of different durations were performed by each cyclist. The power output achieved during the all-out trials was compared with the predicted power output by the APR model. The power output predicted by the model showed very large to nearly perfect correlations to the actual power output obtained during the all-out trials for each cyclist (r = .88 ± .21, .92 ± .17, .95 ± .13, and .97 ± .09). Power output during the all-out trials remained within an average of 6.6% (53 W) of the predicted power output by the model. This preliminary pilot study presents 4 case studies on the applicability of the APR model in professional cyclists using a field-based approach. The decrement in all-out performance during high-intensity exercise seems to conform to a general relationship with a single exponential-decay model describing the decrement in power vs increasing duration. These results are in line with previous studies using the APR model to predict performance during brief all-out trials. Future research should evaluate the APR model with a larger sample size of elite cyclists.

  19. 2 CFR 1400.10 - What does this part do?

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 2 Grants and Agreements 1 2011-01-01 2011-01-01 false What does this part do? 1400.10 Section 1400.10 Grants and Agreements Federal Agency Regulations for Grants and Agreements DEPARTMENT OF THE... Office of Management and Budget (OMB) guidance in subparts A through I of 2 CFR part 180, as supplemented...

  20. 21 CFR 177.1400 - Hydroxyethyl cellulose film, water-insoluble.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 3 2011-04-01 2011-04-01 false Hydroxyethyl cellulose film, water-insoluble. 177.1400 Section 177.1400 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN CONSUMPTION (CONTINUED) INDIRECT FOOD ADDITIVES: POLYMERS Substances for Use as Basic Components of Single and...

  1. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

  2. 7 CFR 1400.203 - Joint operations.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... OF AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.203 Joint operations. (a) A...

  3. 40 CFR 60.1400 - What must I include in my initial report?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... supporting calculations as specified in § 60.1370(a)(1) and (2). (g) If you choose to monitor carbon dioxide... report? 60.1400 Section 60.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... Modification or Reconstruction is Commenced After June 6, 2001 Reporting § 60.1400 What must I include in my...

  4. 40 CFR 60.1400 - What must I include in my initial report?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... supporting calculations as specified in § 60.1370(a)(1) and (2). (g) If you choose to monitor carbon dioxide... report? 60.1400 Section 60.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... Modification or Reconstruction is Commenced After June 6, 2001 Reporting § 60.1400 What must I include in my...

  5. 40 CFR 60.1400 - What must I include in my initial report?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... supporting calculations as specified in § 60.1370(a)(1) and (2). (g) If you choose to monitor carbon dioxide... report? 60.1400 Section 60.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... Modification or Reconstruction is Commenced After June 6, 2001 Reporting § 60.1400 What must I include in my...

  6. 40 CFR 60.1400 - What must I include in my initial report?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... supporting calculations as specified in § 60.1370(a)(1) and (2). (g) If you choose to monitor carbon dioxide... report? 60.1400 Section 60.1400 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... Modification or Reconstruction is Commenced After June 6, 2001 Reporting § 60.1400 What must I include in my...

  7. The Rockwell SR-100G reactor turboelectric space power system

    NASA Technical Reports Server (NTRS)

    Anderson, R. V.

    1985-01-01

    During FY 1982 and 1983, Rockwell International performed system and subsystem studies for space reactor power systems. These studies drew on the expertise gained from the design and flight of the SNAP-10A space nuclear reactor system. These studies, performed for the SP-100 Program, culminated in the selection of a reactor-turboelectric (gas Brayton) system for the SP-100 application; this system is called the SR-100G. This paper describes the features of the system and provides references where more detailed information can be obtained.

  8. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

  9. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

  10. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

  11. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

  12. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

  13. 15 CFR 1400.2 - Definitions.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... as members of a group without regard to their individual qualities. (c) Economically disadvantaged... to diminished capital and credit opportunities because of their identity as members of a group... DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.2 Definitions. For the...

  14. 15 CFR 1400.2 - Definitions.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... as members of a group without regard to their individual qualities. (c) Economically disadvantaged... to diminished capital and credit opportunities because of their identity as members of a group... DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.2 Definitions. For the...

  15. 15 CFR 1400.2 - Definitions.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... as members of a group without regard to their individual qualities. (c) Economically disadvantaged... to diminished capital and credit opportunities because of their identity as members of a group... DEVELOPMENT AGENCY DETERMINATION OF GROUP ELIGIBILITY FOR MBDA ASSISTANCE § 1400.2 Definitions. For the...

  16. Tokamak power reactor ignition and time dependent fractional power operation

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

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transportmore » power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.« less

  17. 7 CFR 1400.106 - Payment limits.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.106 Payment limits. (a) Payments made to...

  18. Lunar in-core thermionic nuclear reactor power system conceptual design

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

    1991-01-01

    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  19. APR3 modulates oxidative stress and mitochondrial function in ARPE-19 cells.

    PubMed

    Li, Yuan; Zou, Xuan; Gao, Jing; Cao, Ke; Feng, Zhihui; Liu, Jiankang

    2018-05-24

    Impairment of retinal pigment epithelial (RPE) cells is considered a key contributor to the development of age-related macular degeneration. Apoptosis-related protein 3 (APR3) was recently discovered after treatment with all- trans retinoic acid, a pivotal molecule in RPE cells. However, the function of APR3 remains poorly understood. In the present study, we found that APR3 could interact with nuclear factor (erythroid-derived 2)-like 2, which is a regulator of phase II enzymes, and that knockdown of APR3 promoted nuclear factor (erythroid-derived 2)-like 2 nuclear translocation and activated expression of phase II enzymes, which was accompanied by improved redox status and mitochondrial activity. Overexpression of APR3 revealed its mitochondrial localization and induced a robust production of reactive oxygen species that was accompanied by impaired mitochondrial oxygen consumption, complex activity, and lower ATP content, resulting in significant changes in mitochondrial structure, which may contribute to cell apoptosis. High doses of all- trans retinoic acid treatment were found to significantly induce APR3 expression, increase reactive oxygen species levels, and decrease ATP content, which were abolished by knockdown of APR3. These results indicate that APR3 plays a vital role in regulating redox status and mitochondrial activity and thus suggest APR3 might be a potential novel target for study of treatment of age-related macular degeneration.-Li, Y., Zou, X., Gao, J., Cao, K., Feng, Z., Liu, J. APR3 modulates oxidative stress and mitochondrial function in ARPE-19 cells.

  20. 7 CFR 1400.503 - Commensurate reduction.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Average Adjusted Gross Income Limitation § 1400.503... legal entity determined to have an average adjusted gross income in excess of the applicable limitation...

  1. 7 CFR 1400.503 - Commensurate reduction.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Average Adjusted Gross Income Limitation § 1400.503... legal entity determined to have an average adjusted gross income in excess of the applicable limitation...

  2. Reactor Power for Large Displacement Autonomous Underwater Vehicles

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

    McClure, Patrick Ray; Reid, Robert Stowers; Poston, David Irvin

    This is a PentaChart on reactor power for large displacement autonomous underwater vehicles. Currently AUVs use batteries or combinations of batteries and fuel cells for power. Battery/fuel cell technology is limited by duration. Batteries and cell fuels are a good match for some missions, but other missions could benefit greatly by a longer duration. The goal is the following: to design nuclear systems to power an AUV and meet design constraints including non-proliferation issues, power level, size constraints, and power conversion limitations. The action plan is to continue development of a range of systems for terrestrial systems and focus onmore » a system for Titan Moon as alternative to Pu-238 for NASA.« less

  3. Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

    2013-08-01

    Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

  4. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems. M.S. Thesis

    NASA Technical Reports Server (NTRS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

  5. The trauma registry compared to All Patient Refined Diagnosis Groups (APR-DRG).

    PubMed

    Hackworth, Jodi; Askegard-Giesmann, Johanna; Rouse, Thomas; Benneyworth, Brian

    2017-05-01

    Literature has shown there are significant differences between administrative databases and clinical registry data. Our objective was to compare the identification of trauma patients using All Patient Refined Diagnosis Related Groups (APR-DRG) as compared to the Trauma Registry and estimate the effects of those discrepancies on utilization. Admitted pediatric patients from 1/2012-12/2013 were abstracted from the trauma registry. The patients were linked to corresponding administrative data using the Pediatric Health Information System database at a single children's hospital. APR-DRGs referencing trauma were used to identify trauma patients. We compared variables related to utilization and diagnosis to determine the level of agreement between the two datasets. There were 1942 trauma registry patients and 980 administrative records identified with trauma-specific APR-DRG during the study period. Forty-two percent (816/1942) of registry records had an associated trauma-specific APR-DRG; 69% of registry patients requiring ICU care had trauma APR-DRGs; 73% of registry patients with head injuries had trauma APR-DRGs. Only 21% of registry patients requiring surgical management had associated trauma APR-DRGs, and 12.5% of simple fractures had associated trauma APR-DRGs. APR-DRGs appeared to only capture a fraction of the entire trauma population and it tends to be the more severely ill patients. As a result, the administrative data was not able to accurately answer hospital or operating room utilization as well as specific information on diagnosis categories regarding trauma patients. APR-DRG administrative data should not be used as the only data source for evaluating the needs of a trauma program. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  7. High power ring methods and accelerator driven subcritical reactor application

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

    Tahar, Malek Haj

    2016-08-07

    High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g.,more » PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field

  8. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

    NASA Technical Reports Server (NTRS)

    Jefferies, K. S.; Tew, R. C.

    1974-01-01

    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  9. Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

    DOEpatents

    LeVert, Francis E.; Cox, Samson A.

    1978-01-01

    A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor.

  10. High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  11. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  12. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

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

    Bernard, J.A.

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

  13. 7 CFR 1400.501 - Determination of average adjusted gross income.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 7 Agriculture 10 2011-01-01 2011-01-01 false Determination of average adjusted gross income. 1400... PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Average Adjusted Gross Income Limitation § 1400.501 Determination of average adjusted gross income. (a) Except as otherwise provided in...

  14. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

  15. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

  16. 7 CFR 1400.402 - Notification.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... social security number or taxpayer identification number of such a person or legal entity, if known, and... SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Foreign Persons § 1400.402 Notification. (a) Any legal entity... legal entity conducts its farming operation if: (1) Any person, group of persons, legal entity, or group...

  17. PR-EDB: Power Reactor Embrittlement Database - Version 3

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

    Wang, Jy-An John; Subramani, Ranjit

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industrymore » standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access

  18. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  19. An RF-Powered Micro-Reactor for Efficient Extraction and Hydrolysis

    NASA Astrophysics Data System (ADS)

    Scott, V.

    2014-12-01

    An RF sample-processing micro-reactor that was developed as part of potential in situ Exploration Missions to inner- and outer-planetary bodies was designed to utilize aqueous solutions subjected to 60 GHz radiation at 730 mW of input power to extract target organic compounds and molecular and inorganic ions as well as to hydrolyze complex polymeric materials. Successful identification and characterization of these molecules relies on the sample-processing techniques utilized alongside state-of-the-art detection and analysis. For mass and power restrictions put on space exploration missions, smaller and more efficient instruments are highly desirable. The RF micro-reactor potentially offers a simplified alternative to the typical gold-standard extractions that often use solvents, chemicals, and conditions that can vary wildly and depend on the targeted molecules. Instead, this instrument uses a single solvent ­— water — that can be "tuned" under the different experimental conditions, leveraging the operating principles of the Sub-Critical Water Extractor. Proof-of-concept experiments examining the hydrolysis of glycosidic and peptide bonds were successful in demonstrating the RF micro-reactor's capabilities. Progress toward coupling the reactor with a micro-scale sample-handling system enabling slurry delivery has been made and preliminary results on heterogeneous reactions and extractions will be presented.

  20. 100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

    NASA Technical Reports Server (NTRS)

    Harty, Richard B.; Mason, Lee S.

    1992-01-01

    Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

  1. Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

    NASA Technical Reports Server (NTRS)

    Tew, R. C.; Jefferies, K. S.

    1974-01-01

    A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

  2. Influence of power supply on the generation of ozone and degradation of phenol in a surface discharge reactor

    NASA Astrophysics Data System (ADS)

    Zhao, Yan; Shang, Kefeng; Duan, Lijuan; Li, Yue; An, Jiutao; Zhang, Chunyang; Lu, Na; Li, Jie; Wu, Yan

    2013-03-01

    A surface Dielectric Barrier Discharge (DBD) reactor was utilized to degrade phenol in water. Different power supplies applied to the DBD reactor affect the discharge modes, the formation of chemically active species and thus the removal efficiency of pollutants. It is thus important to select an optimized power supply for the DBD reactor. In this paper, the influence of the types of power supplies including alternate current (AC) and bipolar pulsed power supply on the ozone generation in a surface discharge reactor was measured. It was found that compared with bipolar pulsed power supply, higher energy efficiency of O3 generation was obtained when DBD reactor was supplied with 50Hz AC power supply. The highest O3 generation was approximate 4 mg kJ-1 moreover, COD removal efficiency of phenol wastewater reached 52.3% after 3 h treatment under an AC peak voltage of 2.6 kV.

  3. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    NASA Astrophysics Data System (ADS)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  4. High density operation for reactor-relevant power exhaust

    NASA Astrophysics Data System (ADS)

    Wischmeier, M.; ASDEX Upgrade Team; Jet Efda Contributors

    2015-08-01

    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  5. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

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

    1984-08-30

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

  6. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

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

  7. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  8. 77 FR 74697 - Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-17

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on January 17, 2013, Room T-2B1, 11545 Rockville Pike...

  9. 40 CFR 1400.12 - Qualified researchers.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.12 Qualified researchers. The Administrator is authorized to provide OCA information, including facility identification, to qualified researchers pursuant to a system developed and...

  10. 40 CFR 1400.12 - Qualified researchers.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.12 Qualified researchers. The Administrator is authorized to provide OCA information, including facility identification, to qualified researchers pursuant to a system developed and...

  11. 40 CFR 1400.12 - Qualified researchers.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.12 Qualified researchers. The Administrator is authorized to provide OCA information, including facility identification, to qualified researchers pursuant to a system developed and...

  12. 40 CFR 1400.12 - Qualified researchers.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.12 Qualified researchers. The Administrator is authorized to provide OCA information, including facility identification, to qualified researchers pursuant to a system developed and...

  13. Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

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

    Karpov, A. S.

    2013-01-15

    A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

  14. Nuclear reactor power for a space-based radar. SP-100 project

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin

    1986-01-01

    A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

  15. 7 CFR 1400.105 - Attribution of payments.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... OF AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.105 Attribution of payments...

  16. 7 CFR 1400.107 - Notification of interests.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... OF AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Limitation § 1400.107 Notification of interests...

  17. Developing the European Center of Competence on VVER-type nuclear power reactors

    NASA Astrophysics Data System (ADS)

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-09-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for VVER-technology is focused on master's degree programmes. The specifics of a systematic approach to training in the area of VVER-type nuclear power reactors technologies are analysed. This paper discusses enhancement of the training opportunities of the European Center that have arisen from advances in methodology and distance education. With a special attention paid to the European Nuclear Education Network (ENEN), the possibilities of further development of the international cooperation between European countries and educational institutions are examined.

  18. 47 CFR 2.1400 - Application for advance approval under part 73.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    .... 2.1400 Section 2.1400 Telecommunication FEDERAL COMMUNICATIONS COMMISSION GENERAL FREQUENCY... standards specified in part 73 of the Rules. The application must include information to show that the... of the encoded aural and visual baseband and transmitted signals and of the encoding equipment used...

  19. Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

    NASA Astrophysics Data System (ADS)

    Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

    2017-01-01

    An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

  20. Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

    NASA Astrophysics Data System (ADS)

    Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

    2017-12-01

    A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

  1. 40 CFR 1400.7 - In general.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.7 In general. The Administrator shall provide OCA information to government officials as provided in this subpart. Any OCA...

  2. 40 CFR 1400.7 - In general.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.7 In general. The Administrator shall provide OCA information to government officials as provided in this subpart. Any OCA...

  3. 40 CFR 1400.7 - In general.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.7 In general. The Administrator shall provide OCA information to government officials as provided in this subpart. Any OCA...

  4. 40 CFR 1400.7 - In general.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.7 In general. The Administrator shall provide OCA information to government officials as provided in this subpart. Any OCA...

  5. Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

    ERIC Educational Resources Information Center

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-01-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

  6. 40 CFR 1400.5 - Internet access to certain off-site consequence analysis data elements.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.5 Internet access to certain off... consequence analysis data elements. 1400.5 Section 1400.5 Protection of Environment ENVIRONMENTAL PROTECTION...

  7. 40 CFR 1400.5 - Internet access to certain off-site consequence analysis data elements.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.5 Internet access to certain off... consequence analysis data elements. 1400.5 Section 1400.5 Protection of Environment ENVIRONMENTAL PROTECTION...

  8. 40 CFR 1400.5 - Internet access to certain off-site consequence analysis data elements.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... UNDER THE CLEAN AIR ACT SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.5 Internet access to certain off... consequence analysis data elements. 1400.5 Section 1400.5 Protection of Environment ENVIRONMENTAL PROTECTION...

  9. Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

    NASA Astrophysics Data System (ADS)

    Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

    2016-12-01

    The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

  10. Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

    Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

    The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

  11. SNAP (Space Nuclear Auxiliary Power) Reactor Overview

    DTIC Science & Technology

    1984-08-01

    so that emphasis could be placed on the development of the space shuttle and the national space station . During 1969 NASA came up with a requirement...which would need the Zr-H reactor system which was the semipermanent orbiting space station . This helped the Zr-H system weather through the major FY 71...provide power for advanced space missions, such as lunar stations or orbiting space platforms, and for interplanetary com- munications. In addition

  12. 76 FR 78173 - Options for Developing the Regulatory Basis for Streamlining Non-Power Reactor License Renewal...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-16

    ... Emergency Preparedness AGENCY: Nuclear Regulatory Commission. ACTION: Notice of public meeting. SUMMARY: The... non-power reactor license renewal and non-power reactor emergency preparedness. This meeting is a... potential enhancements to emergency preparedness requirements. This meeting is open to the public. DATES...

  13. 78 FR 37531 - Agency Information Collection Activities; Comment Request; Annual Performance Reporting (APR...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-21

    ...; Comment Request; Annual Performance Reporting (APR) System for NIDRR Grantees (RERCs, RRTCS, FIPs, ARRTs... of Collection: Annual Performance Reporting (APR) System for NIDRR Grantees (RERCs, RRTCS, FIPs... requests an extension of the Annual Performance Reporting (APR) System for NIDRR Grantees (RERCs, RRTCS...

  14. Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

    NASA Technical Reports Server (NTRS)

    Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

    2002-01-01

    Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

  15. High-temperature, high-power-density thermionic energy conversion for space

    NASA Technical Reports Server (NTRS)

    Morris, J. F.

    1977-01-01

    Theoretic converter outputs and efficiencies indicate the need to consider thermionic energy conversion (TEC) with greater power densities and higher temperatures within reasonable limits for space missions. Converter-output power density, voltage, and efficiency as functions of current density were determined for 1400-to-2000 K emitters with 725-to-1000 K collectors. The results encourage utilization of TEC with hotter-than-1650 K emitters and greater-than-6W sq cm outputs to attain better efficiencies, greater voltages, and higher waste-heat-rejection temperatures for multihundred-kilowatt space-power applications. For example, 1800 K, 30 A sq cm TEC operation for NEP compared with the 1650 K, 5 A/sq cm case should allow much lower radiation weights, substantially fewer and/or smaller emitter heat pipes, significantly reduced reactor and shield-related weights, many fewer converters and associated current-collecting bus bars, less power conditioning, and lower transmission losses. Integration of these effects should yield considerably reduced NEP specific weights.

  16. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  17. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

  18. Compatibility of refractory materials for nuclear reactor poison control systems

    NASA Technical Reports Server (NTRS)

    Sinclair, J. H.

    1974-01-01

    Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

  19. Novel, Integrated Reactor/Power Conversion System (LMR-AMTEC)

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

    Dmitry V. Paramonov, Lead Collaborator

    2001-07-31

    The overall objective of NERI Project Number 99-0198 is to assess the technical and economic feasibility, develop engineering solutions and determine a range of potential applications for ''Novel Integrated Reactor/Energy conversion Systems''. The near term goal is the design of a power supply for developing countries in remote locations in a proliferation resistant, reliable and economical way. The heart of the concept is the use of a single loop liquid metal fast reactor (LMR) with conversion of the heat directly into electricity in a Alkali Metal Thermal to Electric Converter (AMTEC). The first year of the project focused on themore » feasibility issues with a long life, high temperature liquid metal-cooled core; selection of the working fluid, core-to-AMTEC coupling scheme and interface parameters; and, energy conversion systems design and performance. Report Number STD-ES-01-0028, Revision 0, dated July 31, 2001, summarizes the work performed by Westinghouse personnel in Year One and report number UNM-ISNPS-3-2000, dated October 2000, summarizes the work performed by the Institute for Space and Nuclear Power Studies at the University of New Mexico in Year One.« less

  20. 40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.3 Public access to paper copies of off-site consequence analysis information. (a) General. The Administrator and the...

  1. 40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.3 Public access to paper copies of off-site consequence analysis information. (a) General. The Administrator and the...

  2. 40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.3 Public access to paper copies of off-site consequence analysis information. (a) General. The Administrator and the...

  3. 40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.3 Public access to paper copies of off-site consequence analysis information. (a) General. The Administrator and the...

  4. Evaluating SPP/APR Improvement Activities

    ERIC Educational Resources Information Center

    National Early Childhood Technical Assistance Center (NECTAC), 2009

    2009-01-01

    This document is intended to assist State Education Agency (SEA) and Lead Agency (LA) staff and technical assistance providers in designing a meaningful evaluation for the State Performance Plan (SPP)/Annual Performance Report (APR) improvement activities. It provides: (1) information about the relevance of evaluation in the context of improvement…

  5. 40 CFR 1400.8 - Access to off-site consequence analysis information by Federal government officials.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... analysis information by Federal government officials. 1400.8 Section 1400.8 Protection of Environment... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.8 Access to off-site consequence analysis information by Federal...

  6. 40 CFR 1400.8 - Access to off-site consequence analysis information by Federal government officials.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... analysis information by Federal government officials. 1400.8 Section 1400.8 Protection of Environment... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.8 Access to off-site consequence analysis information by Federal...

  7. 40 CFR 1400.8 - Access to off-site consequence analysis information by Federal government officials.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... analysis information by Federal government officials. 1400.8 Section 1400.8 Protection of Environment... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.8 Access to off-site consequence analysis information by Federal...

  8. 40 CFR 1400.8 - Access to off-site consequence analysis information by Federal government officials.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... analysis information by Federal government officials. 1400.8 Section 1400.8 Protection of Environment... INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.8 Access to off-site consequence analysis information by Federal...

  9. Xenon-induced power oscillations in a generic small modular reactor

    NASA Astrophysics Data System (ADS)

    Kitcher, Evans Damenortey

    As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a

  10. Experimental power density distribution benchmark in the TRIGA Mark II reactor

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

    Snoj, L.; Stancar, Z.; Radulovic, V.

    2012-07-01

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the fewmore » available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)« less

  11. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  12. 33 CFR 334.1400 - Pacific Ocean, at Barbers Point, Island of Oahu, Hawaii; restricted area.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ..., Island of Oahu, Hawaii; restricted area. 334.1400 Section 334.1400 Navigation and Navigable Waters CORPS... REGULATIONS § 334.1400 Pacific Ocean, at Barbers Point, Island of Oahu, Hawaii; restricted area. (a) The area... the Officer in Charge, Fleet Area Control and Surveillance Facility, Pearl Harbor, Hawaii 96860-7625...

  13. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  14. 7 CFR 1400.5 - Denial of program benefits.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... OF AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS General Provisions § 1400.5 Denial of program benefits...

  15. Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

    NASA Astrophysics Data System (ADS)

    Nur Krisna, Dwita; Su'ud, Zaki

    2017-01-01

    Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

  16. Apparatus and method for closed-loop control of reactor power in minimum time

    DOEpatents

    Bernard, Jr., John A.

    1988-11-01

    Closed-loop control law for altering the power level of nuclear reactors in a safe manner and without overshoot and in minimum time. Apparatus is provided for moving a fast-acting control element such as a control rod or a control drum for altering the nuclear reactor power level. A computer computes at short time intervals either the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e '.rho.-.SIGMA..beta..sub.i (.lambda..sub.i -.lambda..sub.e ')+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e '.omega.] or the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e .rho.-(.lambda..sub.e /.lambda..sub.e)(.beta.-.rho.)+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e .omega.-(.lambda..sub.e /.lambda..sub.e).omega.] These functions each specify the rate of change of reactivity that is necessary to achieve a specified rate of change of reactor power. The direction and speed of motion of the control element is altered so as to provide the rate of reactivity change calculated using either or both of these functions thereby resulting in the attainment of a new power level without overshoot and in minimum time. These functions are computed at intervals of approximately 0.01-1.0 seconds depending on the specific application.

  17. Space reactor power 1986 - A year of choices and transition

    NASA Technical Reports Server (NTRS)

    Wiley, R. L.; Verga, R. L.; Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.

    1986-01-01

    Both the SP-100 and Multimegawatt programs have made significant progress over the last year and that progress is the focus of this paper. In the SP-100 program the thermoelectric energy conversion concept powered by a compact, high-temperature, lithium-cooled, uranium-nitride-fueled fast spectrum reactor was selected for engineering development and ground demonstration testing at an electrical power level of 300 kilowatts. In the Multimegawatt program, activities moved from the planning phase into one of technology development and assessment with attendant preliminary definition and evaluation of power concepts against requirements of the Strategic Defense Initiative.

  18. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    NASA Astrophysics Data System (ADS)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

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

    NASA Technical Reports Server (NTRS)

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

    1971-01-01

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

  20. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    NASA Astrophysics Data System (ADS)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  1. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  2. 29 CFR 1400.735-20 - Code of Professional Conduct for Labor Mediators.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 4 2010-07-01 2010-07-01 false Code of Professional Conduct for Labor Mediators. 1400.735... Conduct and Responsibilities § 1400.735-20 Code of Professional Conduct for Labor Mediators. In 1964, a Code of Professional Conduct for Labor Mediators was drafted by a Federal-State Liaison Committee and...

  3. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  4. Nuclear reactor power as applied to a space-based radar mission

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

    1988-01-01

    The SP-100 Project was established to develop and demonstrate feasibility of a space reactor power system (SRPS) at power levels of 10's of kilowatts to a megawatt. To help determine systems requirements for the SRPS, a mission and spacecraft were examined which utilize this power system for a space-based radar to observe moving objects. Aspects of the mission and spacecraft bearing on the power system were the primary objectives of this study; performance of the radar itself was not within the scope. The study was carried out by the Systems Design Audit Team of the SP-100 Project.

  5. 13 CFR 107.1400 - Dividends or partnership distributions on 4 percent Preferred Securities.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Dividends or partnership distributions on 4 percent Preferred Securities. 107.1400 Section 107.1400 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION SMALL BUSINESS INVESTMENT COMPANIES SBA Financial Assistance for Licensees...

  6. 40 CFR 1400.11 - Limitation on dissemination to State and local government officials.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Limitation on dissemination to State and local government officials. 1400.11 Section 1400.11 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT...

  7. 7 CFR 1400.502 - Compliance and enforcement.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Average Adjusted Gross Income Limitation § 1400.502 Compliance and enforcement. (a) To comply with the average adjusted gross income limitation, a person or... CCC from a certified public accountant or attorney that the average adjusted gross income of the...

  8. 40 CFR 1400.6 - Enhanced local access.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.6 Enhanced local access. (a) OCA data elements. Consistent with 42 U.S.C...) OCA information. (1) LEPCs and related local government agencies are authorized and encouraged to...

  9. 40 CFR 1400.6 - Enhanced local access.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.6 Enhanced local access. (a) OCA data elements. Consistent with 42 U.S.C...) OCA information. (1) LEPCs and related local government agencies are authorized and encouraged to...

  10. 40 CFR 1400.6 - Enhanced local access.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.6 Enhanced local access. (a) OCA data elements. Consistent with 42 U.S.C...) OCA information. (1) LEPCs and related local government agencies are authorized and encouraged to...

  11. 40 CFR 1400.6 - Enhanced local access.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ...); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Public Access § 1400.6 Enhanced local access. (a) OCA data elements. Consistent with 42 U.S.C...) OCA information. (1) LEPCs and related local government agencies are authorized and encouraged to...

  12. Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

    NASA Technical Reports Server (NTRS)

    Mayo, W.; Lantz, E.

    1973-01-01

    A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

  13. Approach to developing reliable space reactor power systems

    NASA Technical Reports Server (NTRS)

    Mondt, Jack F.; Shinbrot, Charles H.

    1991-01-01

    During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top-down systems approach which includes a point design based on a detailed technical specification of a 100-kW power system. The SP-100 system requirements implicitly recognize the challenge of achieving a high system reliability for a ten-year lifetime, while at the same time using technologies that require very significant development efforts. A low-cost method for assessing reliability, based on an understanding of fundamental failure mechanisms and design margins for specific failure mechanisms, is being developed as part of the SP-100 Program.

  14. Exercise Limitation Imposed by an Approved Air Purifying Respirator (APR)

    DTIC Science & Technology

    2010-05-01

    mentioned that they did not have enough time to inhale, that inspiratory muscles were fatigued , that they got out of rhythm with their breathing and...with APR decreased with time during APR use, a decrease suggesting fatigue of respiratory muscles . Another two subjects did not continue long enough... muscle fatigue , the difference they noted may have been caused by the increase in VE during endurance exercise. The other investigators compared at

  15. Materials technology for an advanced space power nuclear reactor concept: Program summary

    NASA Technical Reports Server (NTRS)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  16. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Van Hagan, Tom H.

    1993-01-01

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

  17. Fast-spectrum space-power-reactor concepts using boron control devices

    NASA Technical Reports Server (NTRS)

    Mayo, W.

    1973-01-01

    Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

  18. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    NASA Astrophysics Data System (ADS)

    Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

    2017-09-01

    Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  19. Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

    NASA Astrophysics Data System (ADS)

    Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

    2017-03-01

    The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

  20. Reactor/Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  1. Fluoride salts as phase change materials for thermal energy storage in the temperature range 1000-1400 K

    NASA Technical Reports Server (NTRS)

    Misra, Ajay K.

    1988-01-01

    Eutectic compositions and congruently melting intermediate compounds in binary and ternary fluoride salt systems were characterized for potential use as latent heat of fusion phase change materials to store thermal energy in the temperature range 1000-1400 K. The melting points and eutectic compositions for many systems with published phase diagrams were experimentally verified and new eutectic compositions having melting points between 1000 and 1400 K were identified. Heats of fusion of several binary and ternary eutectics and congruently melting compounds were experimentally measured by differential scanning calorimetry. For a few systems in which heats of mixing in the melts have been measured, heats of fusion of the eutectics were calculated from thermodynamic considerations and good agreement was obtained between the measured and calculated values. Several combinations of salts with high heats of fusion per unit mass (greater than 0.7 kJ/g) have been identified for possible use as phase change materials in advanced solar dynamic space power applications.

  2. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    NASA Technical Reports Server (NTRS)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  3. 48 CFR 1552.235-76 - Treatment of Confidential Business Information (APR 1996).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Business Information (APR 1996). 1552.235-76 Section 1552.235-76 Federal Acquisition Regulations System... Provisions and Clauses 1552.235-76 Treatment of Confidential Business Information (APR 1996). As prescribed in 1535.007-70(c), insert the following clause: Treatment of Confidential Business Information (TSCA...

  4. A Multiwavelength Study of Nearby Millisecond Pulsar PSR J1400-1431: Improved Astrometry and an Optical Detection of Its Cool White Dwarf Companion

    NASA Astrophysics Data System (ADS)

    Swiggum, J. K.; Kaplan, D. L.; McLaughlin, M. A.; Lorimer, D. R.; Bogdanov, S.; Ray, P. S.; Lynch, R.; Gentile, P.; Rosen, R.; Heatherly, S. A.; Barlow, B. N.; Hegedus, R. J.; Vasquez Soto, A.; Clancy, P.; Kondratiev, V. I.; Stovall, K.; Istrate, A.; Penprase, B.; Bellm, E. C.

    2017-09-01

    In 2012, five high-school students involved in the Pulsar Search Collaboratory discovered the millisecond pulsar (MSP) PSR J1400-1431, and initial timing parameters were published in Rosen et al. a year later. Since then, we have obtained a phase-connected timing solution spanning five years, resolving a significant position discrepancy and measuring \\dot{P}, proper motion, parallax, and a monotonic slope in dispersion measure over time. Due to PSR J1400-1431’s proximity and significant proper motion, we use the Shklovskii effect and other priors to determine a 95% confidence interval for PSR J1400-1431’s distance, d={270}-80+130 pc. With an improved timing position, we present the first detection of the pulsar’s low-mass white dwarf (WD) companion using the Goodman Spectrograph on the 4.1 m SOAR telescope. Deeper imaging suggests that it is a cool DA-type WD with {T}{eff}=3000+/- 100 K and R/{R}⊙ =(2.19+/- 0.03)× {10}-2 (d/270 {pc}). We show a convincing association between PSR J1400-1431 and a γ-ray point source, 3FGL J1400.5-1437, but only weak (3.3σ) evidence of pulsations after folding γ-ray photons using our radio timing model. We detect an X-ray counterpart with XMM-Newton, but the measured X-ray luminosity (1×1029 erg s-1) makes PSR J1400-1431 the least X-ray luminous rotation-powered MSP detected to date. Together, our findings present a consistent picture of a nearby (d≈ 230 pc) MSP in a 9.5-day orbit around a cool ˜0.3 M ⊙ WD companion, with orbital inclination I≳ 60^\\circ .

  5. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

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

    M. L. Grossbeck J-P.A. Renier Tim Bigelow

    2003-09-30

    Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility andmore » cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.« less

  6. Method for producing H.sub.2 using a rotating drum reactor with a pulse jet heat source

    DOEpatents

    Paulson, Leland E.

    1990-01-01

    A method of producing hydrogen by an endothermic steam-carbon reaction using a rotating drum reactor and a pulse jet combustor. The pulse jet combustor uses coal dust as a fuel to provide reaction temperatures of 1300.degree. to 1400.degree. F. Low-rank coal, water, limestone and catalyst are fed into the drum reactor where they are heated, tumbled and reacted. Part of the reaction product from the rotating drum reactor is hydrogen which can be utilized in suitable devices.

  7. Method of production H/sub 2/ using a rotating drum reactor with a pulse jet heat source

    DOEpatents

    Paulson, L.E.

    1988-05-13

    A method of producing hydrogen by an endothermic steam-carbon reaction using a rotating drum reactor and a pulse jet combustor. The pulse jet combustor uses coal dust as a fuel to provide reaction temperatures of 1300/degree/ to 1400/degree/F. Low-rank coal, water, limestone and catalyst are fed into the drum reactor where they are heated, tumbled and reacted. Part of the reaction product from the rotating drum reactor is hydrogen which can be utilized in suitable devices. 1 fig.

  8. Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems

    NASA Technical Reports Server (NTRS)

    Harty, Richard B.; Durand, Richard E.

    1993-01-01

    An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage.

  9. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... analysis information by State and local government officials. 1400.9 Section 1400.9 Protection of... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis...

  10. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... analysis information by State and local government officials. 1400.9 Section 1400.9 Protection of... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis...

  11. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... analysis information by State and local government officials. 1400.9 Section 1400.9 Protection of... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis...

  12. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... analysis information by State and local government officials. 1400.9 Section 1400.9 Protection of... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis...

  13. 40 CFR 1400.9 - Access to off-site consequence analysis information by State and local government officials.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... analysis information by State and local government officials. 1400.9 Section 1400.9 Protection of... CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Access to Off-Site Consequence Analysis Information by Government Officials. § 1400.9 Access to off-site consequence analysis...

  14. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    NASA Astrophysics Data System (ADS)

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-01

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

  15. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

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

    Sheryl Morton; Carl Baily; Tom Hill

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

  16. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

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

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

  17. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    NASA Astrophysics Data System (ADS)

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  18. Multiplex PCR to detect bacteriophages from natural whey cultures of buffalo milk and characterisation of two phages active against Lactococcus lactis, ΦApr-1 and ΦApr-2.

    PubMed

    Aprea, Giuseppe; Mullan, William Michael; Murru, Nicoletta; Fitzgerald, Gerald; Buonanno, Marialuisa; Cortesi, Maria Luisa; Prencipe, Vincenza Annunziata; Migliorati, Giacomo

    2017-09-30

    This work investigated bacteriophage induced starter failures in artisanal buffalo Mozzarella production plants in Southern Italy. Two hundred and ten samples of whey starter cultures were screened for bacteriophage infection. Multiplex polymerase chain reaction (PCR) revealed phage infection in 28.56% of samples, all showing acidification problems during cheese making. Based on DNA sequences, bacteriophages for Lactococcus lactis (L. lactis), Lactobacillus delbruekii (L. delbruekii) and Streptococcus thermophilus (S. thermophilus) were detected. Two phages active against L. lactis, ΦApr-1 and ΦApr-2, were isolated and characterised. The genomes, approximately 31.4 kb and 31 kb for ΦApr-1 and ΦApr-2 respectively, consisted of double-stranded linear DNA with pac-type system. Sodium dodecyl sulfate polyacrylamide gel electrophoresis (SDS‑PAGE) showed one major structural protein of approximately 32.5 kDa and several minor proteins. This is the first report of phage isolation in buffalo milk and of the use of multiplex PCR to screen and study the diversity of phages against Lactic Acid Bacteria (LAB) strains in artisanal Water Buffalo Mozzarella starters.

  19. 14C content in vegetation in the vicinities of Brazilian nuclear power reactors.

    PubMed

    Dias, Cíntia Melazo; Santos, Roberto Ventura; Stenström, Kristina; Nícoli, Iêda Gomes; Skog, Göran; da Silveira Corrêa, Rosangela

    2008-07-01

    (14)C specific activities were measured in grass samples collected around Brazilian nuclear power reactors. The specific activity values varied between 227 and 299 Bq/kg C. Except for two samples which showed (14)C specific activities 22% above background values, half of the samples showed background specific activities, and the other half had a (14)C excess of 1-18%. The highest specific activities were found close to the nuclear power plants and along the main wind directions (NE and NNE). The activity values were found to decrease with increasing distance from the reactors. The unexpectedly high (14)C excess values found in two samples were related to the local topography, which favors (14)C accumulation and limits the dispersion of the plume. The results indicate a clear (14)C anthropogenic signal within 5 km around the nuclear power plants which is most prominent along northeastwards, the prevailing wind direction.

  20. Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

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

    Wood, Richard Thomas

    2008-01-01

    In recent years, the National Aeronautics and Space Administration (NASA) has been considering deep space missions that utilize a small-reactor power system (SRPS) to provide energy for propulsion and spacecraft power. Additionally, application of SRPS modules as a planetary power source is being investigated to enable a continuous human presence for nonpolar lunar sites and on Mars. A SRPS can supply high-sustained power for space and surface applications that is both reliable and mass efficient. The use of small nuclear reactors for deep space or planetary missions presents some unique challenges regarding the operations and control of the power system.more » Current-generation terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a SRPS employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. While surface power systems for planetary outposts face less extreme delays and periods of isolation and may benefit from limited maintenance capabilities, considerations such as human safety, resource limitations and usage priorities, and economics favor minimizing direct, continuous human interaction with the SRPS for online, dedicated power system management. Thus, a SRPS control system for space or planetary missions must provide capabilities for operational autonomy. For terrestrial reactors, large-scale power plants remain the preferred near-term option for nuclear power generation. However, the desire to reduce reliance on carbon-emitting power sources in developing countries may lead to increased consideration of SRPS modules for local power generation in remote regions that are characterized by emerging, less established

  1. Neutron dose estimation in a zero power nuclear reactor

    NASA Astrophysics Data System (ADS)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  2. Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

    NASA Astrophysics Data System (ADS)

    Joung Lim, Mi; Maeng, Young Jae; Fero, Arnold H.; Anderson, Stanwood L.

    2016-02-01

    The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

  3. 40 CFR 1400.10 - Limitation on public dissemination.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.10 Limitation on public dissemination. Except as... officials, and qualified researchers are prohibited from disseminating OCA information and OCA rankings to...

  4. 40 CFR 1400.10 - Limitation on public dissemination.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.10 Limitation on public dissemination. Except as... officials, and qualified researchers are prohibited from disseminating OCA information and OCA rankings to...

  5. 40 CFR 1400.10 - Limitation on public dissemination.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.10 Limitation on public dissemination. Except as... officials, and qualified researchers are prohibited from disseminating OCA information and OCA rankings to...

  6. 40 CFR 1400.10 - Limitation on public dissemination.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... SECTION 112(r)(7); DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION DISTRIBUTION OF OFF-SITE CONSEQUENCE ANALYSIS INFORMATION Other Provisions § 1400.10 Limitation on public dissemination. Except as... officials, and qualified researchers are prohibited from disseminating OCA information and OCA rankings to...

  7. Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

    NASA Astrophysics Data System (ADS)

    Zaman, Badrus; Wardhana, Irawan Wisnu

    2018-02-01

    Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

  8. 48 CFR 1552.235-75 - Access to Toxic Substances Control Act Confidential Business Information (APR 1996).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Control Act Confidential Business Information (APR 1996). 1552.235-75 Section 1552.235-75 Federal... Confidential Business Information (APR 1996). As prescribed in 1535.007(b), insert the following provision: Access to Toxic Substances Control Act Confidential Business Information (APR 1996) In order to perform...

  9. 42 CFR § 414.1400 - Third party data submission.

    Code of Federal Regulations, 2010 CFR

    2017-10-01

    ... SERVICES (CONTINUED) MEDICARE PROGRAM (CONTINUED) PAYMENT FOR PART B MEDICAL AND OTHER HEALTH SERVICES Merit-Based Incentive Payment System and Alternative Payment Model Incentive § 414.1400 Third party data...

  10. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  11. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

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

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

  12. Delayed Neutrons Effect on Power Reactor with Variation of Fluid Fuel Velocity at MSR Fuji-12

    NASA Astrophysics Data System (ADS)

    Kuncoro Aji, Indarta; Pramuditya, Syeilendra; Novitrian; Irwanto, Dwi; Waris, Abdul

    2017-01-01

    As the nuclear reactor operate with liquid fuel, controlling velocity of the fuel flow on the Molten salt reactor very influence on the neutron kinetics in that reactor system. The effect of the pace fuel changes to the populations number of neutrons and power density on vertical direction (1 dimension) from the first until fifth year reactor operating had been analyzed on this research. This research had been conducted on MSR Fuji-12 with a two meters core high, and LiF-BeF2-ThF4-233UF4 as fuel composition respectively 71.78%-16%-11.86%-0.36%. Data of reactivity, neutron flux, and the macroscopic fission cross section obtained from ouput of SRAC (neutronic calculation code has been developed by JAEA, with JENDL-4.0 as data library on the SRAC calculation) was being used for the calculation process of this research. The calculation process of this research had been performed numerically by SOR (successive over relaxation) and finite difference methode, as well as using C programing language. From the calculation, regarding to the value of power density resulting from delayed neutrons, concluded that 20 m/s is the optimum fuel flow velocity in all the years reactor had operated. Where the increases number of power are inversely proportional with the fuel flow speed.

  13. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-01

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the

  14. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

  15. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

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

  16. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

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

  17. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

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

  18. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  19. 7 CFR 1400.210 - Deceased and incapacitated persons.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... CORPORATION, DEPARTMENT OF AGRICULTURE GENERAL REGULATIONS AND POLICIES PAYMENT LIMITATION AND PAYMENT ELIGIBILITY FOR 2009 AND SUBSEQUENT CROP, PROGRAM, OR FISCAL YEARS Payment Eligibility § 1400.210 Deceased and... determination to be in effect for that program year or fiscal year, as applicable. However, the following year...

  20. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2009-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  1. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  2. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

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

    Sterbentz, James William; Werner, James Elmer; McKellar, Michael George

    The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heatmore » exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.« less

  3. 40 CFR 1400.3 - Public access to paper copies of off-site consequence analysis information.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Public access to paper copies of off-site consequence analysis information. 1400.3 Section 1400.3 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY AND DEPARTMENT OF JUSTICE ACCIDENTAL RELEASE PREVENTION REQUIREMENTS; RISK MANAGEMENT...

  4. Development of Nanosatellite Technology with APRS Module for Disaster Mitigation

    NASA Astrophysics Data System (ADS)

    Prahyang, S. Y.; Dhiya’Ulhaq, M. Z.; Golim, O. P.; Gunawan, R.; Suhandinata; Jahja, E.; Nelwan, E. R. G.; Ananta, C.; Chow, I. M.; Mali, N. D. F.

    2018-05-01

    Development of nanosatellite technology has enabled satellites to be developed with multiple capabilities for a specific mission in a short time with a low cost. Satellite communications are proved to be more effective in delivering information due to its large coverage area. Surya Satellite-1 will become the first Indonesian nanosatellite developed by undergraduate students. It is designed with low-cost commercial payloads, including an APRS module for communication and operated on VHF and UHF amateur radio frequencies. The mission of the satellites focused on disaster mitigation through APRS communication network with remote stations located on disaster-prone areas.

  5. 29 CFR 1400.735-21 - Miscellaneous statutory provisions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 29 Labor 4 2010-07-01 2010-07-01 false Miscellaneous statutory provisions. 1400.735-21 Section... Service.” (b) Chapter 11 of title 18, United States Code, relating to bribery, graft, and conflicts of... Registration Act (18 U.S.C. 219). (r) Penalties: The following table, copied from the Federal Personnel Manual...

  6. A high temperature drop-tube and packed-bed solar reactor for continuous biomass gasification

    NASA Astrophysics Data System (ADS)

    Bellouard, Quentin; Abanades, Stéphane; Rodat, Sylvain; Dupassieux, Nathalie

    2017-06-01

    Biomass gasification is an attractive process to produce high-value syngas. Utilization of concentrated solar energy as the heat source for driving reactions increases the energy conversion efficiency, saves biomass resource, and eliminates the needs for gas cleaning and separation. A high-temperature tubular solar reactor combining drop tube and packed bed concepts was used for continuous solar-driven gasification of biomass. This 1 kW reactor was experimentally tested with biomass feeding under real solar irradiation conditions at the focus of a 2 m-diameter parabolic solar concentrator. Experiments were conducted at temperatures ranging from 1000°C to 1400°C using wood composed of a mix of pine and spruce (bark included) as biomass feedstock. The aim of this study was to demonstrate the feasibility of syngas production in this reactor concept and to prove the reliability of continuous biomass gasification processing using solar energy. The study first consisted of a parametric study of the gasification conditions to obtain an optimal gas yield. The influence of temperature and oxidizing agent (H2O or CO2) on the product gas composition was investigated. The study then focused on solar gasification during continuous biomass particle injection for demonstrating the feasibility of a continuous process. Regarding the energy conversion efficiency of the lab scale reactor, energy upgrade factor of 1.21 and solar-to-fuel thermochemical efficiency up to 28% were achieved using wood heated up to 1400°C.

  7. Power control of SAFE reactor using fuzzy logic

    NASA Astrophysics Data System (ADS)

    Irvine, Claude

    2002-01-01

    Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc. .

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

    NASA Astrophysics Data System (ADS)

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

    2016-03-01

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

  9. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    NASA Technical Reports Server (NTRS)

    Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

    2011-01-01

    The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

  10. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    NASA Astrophysics Data System (ADS)

    Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

    The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

  11. Removal of toxic uranium from synthetic nuclear power reactor effluents using uranyl ion imprinted polymer particles.

    PubMed

    Preetha, Chandrika Ravindran; Gladis, Joseph Mary; Rao, Talasila Prasada; Venkateswaran, Gopala

    2006-05-01

    Major quantities of uranium find use as nuclear fuel in nuclear power reactors. In view of the extreme toxicity of uranium and consequent stringent limits fixed by WHO and various national governments, it is essential to remove uranium from nuclear power reactor effluents before discharge into environment. Ion imprinted polymer (IIP) materials have traditionally been used for the recovery of uranium from dilute aqueous solutions prior to detection or from seawater. We now describe the use of IIP materials for selective removal of uranium from a typical synthetic nuclear power reactor effluent. The IIP materials were prepared for uranyl ion (imprint ion) by forming binary salicylaldoxime (SALO) or 4-vinylpyridine (VP) or ternary SALO-VP complexes in 2-methoxyethanol (porogen) and copolymerizing in the presence of styrene (monomer), divinylbenzene (cross-linking monomer), and 2,2'-azobisisobutyronitrile (initiator). The resulting materials were then ground and sieved to obtain unleached polymer particles. Leached IIP particles were obtained by leaching the imprint ions with 6.0 M HCl. Control polymer particles were also prepared analogously without the imprint ion. The IIP particles obtained with ternary complex alone gave quantitative removal of uranyl ion in the pH range 3.5-5.0 with as low as 0.08 g. The retention capacity of uranyl IIP particles was found to be 98.50 mg/g of polymer. The present study successfully demonstrates the feasibility of removing uranyl ions selectively in the range 5 microg - 300 mg present in 500 mL of synthetic nuclear power reactor effluent containing a host of other inorganic species.

  12. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  13. Nuclear reactor power as applied to a space-based radar mission

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    A space-based radar mission and spacecraft are examined to determine system requirements for a 300 kWe space nuclear reactor power system. The spacecraft configuration and its orbit, launch vehicle, and propulsion are described. Mission profiles are addressed, and storage in assembly orbit is considered. Dynamics and attitude control and the problems of nuclear and thermal radiation are examined.

  14. Effect of inlet temperature on the performance of a catalytic reactor. [air pollution control

    NASA Technical Reports Server (NTRS)

    Anderson, D. N.

    1978-01-01

    A 12 cm diameter by 15 cm long catalytic reactor was tested with No. 2 diesel fuel in a combustion test rig at inlet temperatures of 700, 800, 900, and 1000 K. Other test conditions included pressures of 3 and 6 x 10 to the 5th power Pa, reference velocities of 10, 15, and 20 m/s, and adiabatic combustion temperatures in the range 1100 to 1400 K. The combustion efficiency was calculated from measurements of carbon monoxide and unburned hydrocarbon emissions. Nitrogen oxide emissions and reactor pressure drop were also measured. At a reference velocity of 10 m/s, the CO and unburned hydrocarbons emissions, and, therefore, the combustion efficiency, were independent of inlet temperature. At an inlet temperature of 1000 K, they were independent of reference velocity. Nitrogen oxides emissions resulted from conversion of the small amount (135 ppm) of fuel-bound nitrogen in the fuel. Up to 90 percent conversion was observed with no apparent effect of any of the test variables. For typical gas turbine operating conditions, all three pollutants were below levels which would permit the most stringent proposed automotive emissions standards to be met.

  15. 76 FR 32240 - Advisory Committee on Reactor Safeguards (ACRS) Meeting on the ACRS Subcommittee on Power Uprates

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-03

    ... Expanded Operating Domains-Power Distribution Validation and Pin-by-Pin Gamma Scan). The Subcommittee will... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting on the ACRS Subcommittee on Power Uprates Notice of Meeting The ACRS Subcommittee on Power Uprates will hold a meeting on...

  16. APR-246/PRIMA-1Met Inhibits and Reverses Squamous Metaplasia in Human Conjunctival Epithelium.

    PubMed

    Li, Jing; Li, Cheng; Wang, Guoliang; Liu, Zhen; Chen, Pei; Yang, Qichen; Dong, Nuo; Wu, Huping; Liu, Zuguo; Li, Wei

    2016-02-01

    Squamous metaplasia is a common pathologic condition in ocular surface diseases for which there is no therapeutic medication in clinic. In this study, we investigated the effect of a small molecule, APR-246/PRIMA-1(Met), on squamous metaplasia in human conjunctival epithelium. Human conjunctival explants were cultured for up to 12 days under airlifting conditions. Epithelial cell differentiation and proliferation were assessed by Cytokeratin 10 (K10), K14, K19, Pax6, MUC5AC, and p63 immunostaining patterns. β-catenin and TCF-4 immunofluorescent staining and real-time PCR characterized Wnt signaling pathway involvement. Pterygium clinical samples were cultured under airlifting conditions with or without APR-246 for 4 days. p63, K10, β-catenin, and TCF-4 expression in pterygial epithelium was determined by immunofluorescent staining and real-time PCR. Airlift conjunctival explants resulted in increased stratification and intrastromal epithelial invagination. Such pathology was accompanied by increases in K10, K14, and p63 expression, whereas K19 and Pax6 levels declined when compared to those in freshly isolated tissue. On the other hand, APR-246 reversed all of these declines in K10, K14, and p63 expression. Furthermore, K19 and Pax6 increased along with rises in goblet cell density. These effects of APR-246 were accompanied by near restoration of normal conjunctival epithelial histology. APR-246 also reversed squamous metaplasia in pterygial epithelium that had developed after 4 days in ex vivo culture. Reductions in squamous metaplasia induced by APR-246 suggest it may provide a novel therapeutic approach in different squamous metaplasia-associated ocular surface diseases.

  17. Feasibility study of computed vs measured high b-value (1400 s/mm²) diffusion-weighted MR images of the prostate

    PubMed Central

    Bittencourt, Leonardo K; Attenberger, Ulrike I; Lima, Daniel; Strecker, Ralph; de Oliveira, Andre; Schoenberg, Stefan O; Gasparetto, Emerson L; Hausmann, Daniel

    2014-01-01

    AIM: To evaluate the impact of computed b = 1400 s/mm2 (C-b1400) vs measured b = 1400 s/mm2 (M-b1400) diffusion-weighted images (DWI) on lesion detection rate, image quality and quality of lesion demarcation using a modern 3T-MR system based on a small-field-of-view sequence (sFOV). METHODS: Thirty patients (PSA: 9.5 ± 8.7 ng/mL; 68 ± 12 years) referred for magnetic resonance imaging (MRI) of the prostate were enrolled in this study. All measurements were performed on a 3T MR system. For DWI, a single-shot EPI diffusion sequence (b = 0, 100, 400, 800 s/mm²) was utilized. C-b1400 was calculated voxelwise from the ADC and diffusion images. Additionally, M-b1400 was acquired for evaluation and comparison. Lesion detection rate and maximum lesion diameters were obtained and compared. Image quality and quality of lesion demarcation were rated according to a 5-point Likert-type scale. Ratios of lesion-to-bladder as well as prostate-to-bladder signal intensity (SI) were calculated to estimate the signal-to-noise-ratio (SNR). RESULTS: Twenty-four lesions were detected on M-b1400 images and compared to C-b1400 images. C-b1400 detected three additional cancer suspicious lesions. Overall image quality was rated significantly better and SI ratios were significantly higher on C-b1400 (2.3 ± 0.8 vs 3.1 ± 1.0, P < 0.001; 5.6 ± 1.8 vs 2.8 ± 0.9, P < 0.001). Comparison of lesion size showed no significant differences between C- and M-b1400 (P = 0.22). CONCLUSION: Combination of a high b-value extrapolation and sFOV may contribute to increase diagnostic accuracy of DWI without an increase of acquisition time, which may be useful to guide targeted prostate biopsies and to improve quality of multiparametric MRI (mMRI) especially under economical aspects in a private practice setting. PMID:24976938

  18. PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

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

    Vondy, D.R.

    1979-10-01

    The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

  19. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  20. 77 FR 9707 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Power Uprates...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-17

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Power Uprates; Revision to February 23, 2012, ACRS Meeting Federal Register Notice The Federal Register Notice for the ACRS Subcommittee meeting on Power Uprates, scheduled to be held on February 23...

  1. The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

    NASA Astrophysics Data System (ADS)

    Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

    2013-10-01

    A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

  2. Neutron fluxes in test reactors

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

    Youinou, Gilles Jean-Michel

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  3. Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program

    NASA Technical Reports Server (NTRS)

    Ambrus, J. H.; Wright, W. E.; Bunch, D. F.

    1984-01-01

    The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

  4. 17 CFR 37.1400 - Core Principle 14-System safeguards.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... procedures, and automated systems, that: (1) Are reliable and secure; and (2) Have adequate scalable capacity... 17 Commodity and Securities Exchanges 1 2014-04-01 2014-04-01 false Core Principle 14-System... SWAP EXECUTION FACILITIES System Safeguards § 37.1400 Core Principle 14—System safeguards. The swap...

  5. The Ongoing Rediscovery of Après-Coup as a Central Freudian Concept.

    PubMed

    House, Jonathan

    2017-10-01

    Après-coup, Freud's Nachträglichkeit, is an essential psychoanalytic concept structuring each of four concepts, four mental processes that lie at the foundation of Freud's thinking: psychic trauma, repression, the creation of the unconscious, and the creation of infantile sexuality. It is argued here that infantile sexual drives, in contrast to the self-preservative instincts, arise from a two-step process of translation and repression in which the residues of failed translation become source-objects of the drives. These residues of failed translation have an associative resonance with adult sexuality, and the child is driven to ongoing attempts to translate them, to make them meaningful après coup. Thus, après-coup is at the heart of the human subject as a sexual creature who requires, desires, and creates meaning.

  6. 29 CFR 1400.735-61 - Notice to and appeal of employee.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... STANDARDS OF CONDUCT, RESPONSIBILITIES, AND DISCIPLINE Disciplinary Actions and Penalties § 1400.735-61... disciplinary actions against status employees. Such proceedings will include notification to the employee of...

  7. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

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

    Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to

  8. Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

    NASA Technical Reports Server (NTRS)

    Bowles, K. J.

    1974-01-01

    An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

  9. The performance of hafnium and gadolinium self powered neutron detectors in the TREAT reactor

    NASA Astrophysics Data System (ADS)

    Imel, G. R.; Hart, P. R.

    1996-05-01

    The use of gadolinium and hafnium self powered neutron detectors in a transient reactor is described in this paper. The detectors were calibrated to the fission rate of U-235 using calibrated fission chambers; the calibration factors were tested in two reactors in steady state and found to be consistent. Calibration of the detectors in transient reactor conditions was done by using uranium wires that were analyzed by radiochemistry techniques to determine total fissions during the transient. This was correlated to the time-integrated current of the detectors during the transient. A temperature correction factor was derived to account for self-shielding effects in the hafnium and gadolinium detectors. The dynamic response of the detectors under transient conditions was studied, and found to be excellent.

  10. Inducible NOS inhibitor 1400W reduces hypoxia/re-oxygenation injury in rat lung.

    PubMed

    Rus, Alma; Castro, Lourdes; Del Moral, Maria Luisa; Peinado, Angeles

    2010-01-01

    Nitric oxide (NO(*)) from inducible NO(*) synthase (iNOS) has been reported to either protect against, or contribute to, hypoxia/re-oxygenation lung injury. The present work aimed to clarify this double role in the hypoxic lung. With this objective, a follow-up study was made in Wistar rats submitted to hypoxia/re-oxygenation (hypoxia for 30 min; re-oxygenation of 0 h, 48 h, and 5 days), with or without prior treatment with the selective iNOS inhibitor 1400W (10 mg/kg). NO(*) levels (NOx), lipid peroxidation, apoptosis, and protein nitration were analysed. This is the first time-course study which investigates the effects of 1400W during hypoxia/re-oxygenation in the rat lung. The results showed that the administration of 1400W lowered NOx levels in all the experimental groups. In addition, lipid peroxidation, the percentage of apoptotic cells, and nitrated protein expression fell in the late post-hypoxia period (48 h and 5 days). Our results reveal that the inhibition of iNOS in the hypoxic lung reduced the damage observed before the treatment with 1400W, suggesting that iNOS-derived NO(*) may exert a negative effect on this organ during hypoxia/re-oxygenation. These findings are notable, since they indicate that any therapeutic strategy aimed at controlling excess generation of NO(*) from iNOS may be useful in alleviating NO(*)-mediated adverse effects in hypoxic lungs.

  11. Near-infrared photon time-of-flight spectroscopy of turbid materials up to 1400 nm

    NASA Astrophysics Data System (ADS)

    Svensson, Tomas; Alerstam, Erik; Khoptyar, Dmitry; Johansson, Jonas; Folestad, Staffan; Andersson-Engels, Stefan

    2009-06-01

    Photon time-of-flight spectroscopy (PTOFS) is a powerful tool for analysis of turbid materials. We have constructed a time-of-flight spectrometer based on a supercontinuum fiber laser, acousto-optical tunable filtering, and an InP/InGaAsP microchannel plate photomultiplier tube. The system is capable of performing PTOFS up to 1400 nm, and thus covers an important region for vibrational spectroscopy of solid samples. The development significantly increases the applicability of PTOFS for analysis of chemical content and physical properties of turbid media. The great value of the proposed approach is illustrated by revealing the distinct absorption features of turbid epoxy resin. Promising future applications of the approach are discussed, including quantitative assessment of pharmaceuticals, powder analysis, and calibration-free near-infrared spectroscopy.

  12. Carbon-coated ceramic membrane reactor for the production of hydrogen by aqueous-phase reforming of sorbitol.

    PubMed

    Neira D'Angelo, M F; Ordomsky, V; Schouten, J C; van der Schaaf, J; Nijhuis, T A

    2014-07-01

    Hydrogen was produced by aqueous-phase reforming (APR) of sorbitol in a carbon-on-alumina tubular membrane reactor (4 nm pore size, 7 cm long, 3 mm internal diameter) that allows the hydrogen gas to permeate to the shell side, whereas the liquid remains in the tube side. The hydrophobic nature of the membrane serves to avoid water loss and to minimize the interaction between the ceramic support and water, thus reducing the risks of membrane degradation upon operation. The permeation of hydrogen is dominated by the diffusivity of the hydrogen in water. Thus, higher operation temperatures result in an increase of the flux of hydrogen. The differential pressure has a negative effect on the flux of hydrogen due to the presence of liquid in the larger pores. The membrane was suitable for use in APR, and yielded 2.5 times more hydrogen than a reference reactor (with no membrane). Removal of hydrogen through the membrane assists in the reaction by preventing its consumption in undesired reactions. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

    NASA Astrophysics Data System (ADS)

    Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

    2015-05-01

    The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

  14. Development of an inconel self powered neutron detector for in-core reactor monitoring

    NASA Astrophysics Data System (ADS)

    Alex, M.; Ghodgaonkar, M. D.

    2007-04-01

    The paper describes the development and testing of an Inconel600 (2 mm diameter×21 cm long) self-powered neutron detector for in-core neutron monitoring. The detector has 3.5 mm overall diameter and 22 cm length and is integrally coupled to a 12 m long mineral insulated cable. The performance of the detector was compared with cobalt and platinum detectors of similar dimensions. Gamma sensitivity measurements performed at the 60Co irradiation facility in 14 MR/h gamma field showed values of -4.4×10 -18 A/R/h/cm (-9.3×10 -24 A/ γ/cm 2-s/cm), -5.2×10 -18 A/R/h/cm (-1.133×10 -23 A/ γ/cm 2-s/cm) and 34×10 -18 A/R/h/cm (7.14×10 -23 A/ γ/cm 2-s/cm) for the Inconel, Co and Pt detectors, respectively. The detectors together with a miniature gamma ion chamber and fission chamber were tested in the in-core Apsara Swimming Pool type reactor. The ion chambers were used to estimate the neutron and gamma fields. With an effective neutron cross-section of 4b, the Inconel detector has a total sensitivity of 6×10 -23 A/nv/cm while the corresponding sensitivities for the platinum and cobalt detectors were 1.69×10 -22 and 2.64×10 -22 A/nv/cm. The linearity of the detector responses at power levels ranging from 100 to 200 kW was within ±5%. The response of the detectors to reactor scram showed that the prompt response of the Inconel detector was 0.95 while it was 0.7 and 0.95 for the platinum and cobalt self-powered detectors, respectively. The detector was also installed in the horizontal flux unit of 540 MW Pressurised Heavy Water Reactor (PHWR). The neutron flux at the detector location was calculated by Triveni code. The detector response was measured from 0.02% to 0.07% of full power and showed good correlation between power level and detector signals. Long-term tests and the dynamic response of the detector to shut down in PHWR are in progress.

  15. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  16. 12. Historic American Buildings Survey Arthur C. Haskell, Photographer Apr. ...

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

    12. Historic American Buildings Survey Arthur C. Haskell, Photographer Apr. 1, 1939 (l) INT.- STAIRWAY, 4th FLOOR, LOOKING SOUTH - M.I.T., Rogers Building, 491 Boylston Street, Boston, Suffolk County, MA

  17. 9. Historic American Buildings Survey Arthur C. Haskell, Photographer Apr. ...

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

    9. Historic American Buildings Survey Arthur C. Haskell, Photographer Apr. 1, 1939 (i) INT.- STAIR HALL, 1st. FLOOR, LOOKING NORTH - M.I.T., Rogers Building, 491 Boylston Street, Boston, Suffolk County, MA

  18. A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

    NASA Astrophysics Data System (ADS)

    Bertch, Timothy Creston

    1998-12-01

    Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release

  19. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

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

    REID, ROBERT S.; PEARSON, J. BOSIE; STEWART, ERIC T.

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WSTmore » is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.« less

  20. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

    NASA Technical Reports Server (NTRS)

    Juhasz, A. J.; Jones, B. I.

    1987-01-01

    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  1. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

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

  2. AN EVALUATION OF HEAVY WATER REACTORS FOR POWER

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

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

    1957-10-01

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

  3. Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

    NASA Astrophysics Data System (ADS)

    Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

    2010-07-01

    Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

  4. 15 CFR 1400.4 - Evidence of social or economic disadvantage.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Evidence of social or economic... ASSISTANCE § 1400.4 Evidence of social or economic disadvantage. (a) The representatives of the group requesting formal designation should establish social or economic disadvantage by a preponderance of the...

  5. 15 CFR 1400.4 - Evidence of social or economic disadvantage.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 15 Commerce and Foreign Trade 3 2012-01-01 2012-01-01 false Evidence of social or economic... ASSISTANCE § 1400.4 Evidence of social or economic disadvantage. (a) The representatives of the group requesting formal designation should establish social or economic disadvantage by a preponderance of the...

  6. 15 CFR 1400.4 - Evidence of social or economic disadvantage.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 15 Commerce and Foreign Trade 3 2013-01-01 2013-01-01 false Evidence of social or economic... ASSISTANCE § 1400.4 Evidence of social or economic disadvantage. (a) The representatives of the group requesting formal designation should establish social or economic disadvantage by a preponderance of the...

  7. 15 CFR 1400.4 - Evidence of social or economic disadvantage.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 15 Commerce and Foreign Trade 3 2011-01-01 2011-01-01 false Evidence of social or economic... ASSISTANCE § 1400.4 Evidence of social or economic disadvantage. (a) The representatives of the group requesting formal designation should establish social or economic disadvantage by a preponderance of the...

  8. 15 CFR 1400.4 - Evidence of social or economic disadvantage.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 15 Commerce and Foreign Trade 3 2014-01-01 2014-01-01 false Evidence of social or economic... ASSISTANCE § 1400.4 Evidence of social or economic disadvantage. (a) The representatives of the group requesting formal designation should establish social or economic disadvantage by a preponderance of the...

  9. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. B.; Reid, R.; Sadasivan, P.; Stewart, E.

    2007-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A representative lunar surface reactor design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The evaluation compares the experimental data from the WST to CFD models. Performance of a water shield on the lunar surface is predicted by CFD models anchored to test data, and by matching relevant dimensionless parameters.

  10. Wireless Powered Cooperative Communications: Power-Splitting Relaying With Energy Accumulation (Author’s Manuscript)

    DTIC Science & Technology

    2016-03-21

    2016 2 i.e., wireless power transfer (WPT) and wireless information transfer (WIT), fundamental changes to the designs of green communication networks...simulta- neous wireless information and power transfer ,” IEEE Commun. Mag., vol. 53, no. 4, pp. 86–93, Apr. 2015. [6] H. Tabassum, E. Hossain, A...broadcasting for simultaneous wire- less information and power transfer ,” IEEE Trans. Wireless Commun., vol. 12, no. 5, pp. 1989–2001, May 2013. [9] K. Huang

  11. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Houts, Michael

    2001-02-01

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

  12. Calculation of energetic characteristics of C-14 emitted from Beloyarsk nuclear power plant plume with fast neutron reactor

    NASA Astrophysics Data System (ADS)

    Kolotkov, Gennady A.; Penin, Sergei

    2017-11-01

    The paper examines an update of comparative analysis of radionuclides released into the atmosphere from Beloyarsk nuclear power plant with fast-neutron reactor for nine years in a row, from 2008 to 2016. It has been shown that the main radionuclides throw out into the atmosphere from Beloyarsk nuclear power plant are beta-active radionuclides. Based on data releases of the RPA "Typhoon", it has been conclude that radiation situation become worse insignificantly; beside on the new reactor BN-800 was put in operation in 2016. Using Spencer-Fano's equation, it was carried out the summary spectrum of emitted radionuclides. On example of Beloyarsk nuclear power plant, it was considered a question about ability of remote detection of raised radioactivity in the atmospheric radioactive plume. It has been shown that it possible to detect raised radioactivity in the emission plume from Beloyarsk nuclear power plant.

  13. Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

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

    Palabrica, R.J.

    1981-01-01

    The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

  14. Investigation on discharge characteristics of a coaxial dielectric barrier discharge reactor driven by AC and ns power sources

    NASA Astrophysics Data System (ADS)

    Qian, WANG; Feng, LIU; Chuanrun, MIAO; Bing, YAN; Zhi, FANG

    2018-03-01

    A coaxial dielectric barrier discharge (DBD) reactor with double layer dielectric barriers has been developed for exhaust gas treatment and excited either by AC power or nanosecond (ns) pulse to generate atmospheric pressure plasma. The comparative study on the discharge characteristics of the discharge uniformity, power deposition, energy efficiency, and operation temperature between AC and ns pulsed coaxial DBD is carried out in terms of optical and electrical characteristics and operation temperature for optimizing the coaxial DBD reactor performance. The voltages across the air gap and dielectric layer and the conduction and displacement currents are extracted from the applied voltages and measured currents of AC and ns pulsed coaxial DBDs for the calculation of the power depositions and energy efficiencies through an equivalent electrical model. The discharge uniformity and operating temperature of the coaxial DBD reactor are monitored and analyzed by optical images and infrared camera. A heat conduction model is used to calculate the temperature of the internal quartz tube. It is found that the ns pulsed coaxial DBD has a much higher instantaneous power deposition in plasma, a lower total power consumption, and a higher energy efficiency compared with that excited by AC power and is more homogeneous and stable. The temperature of the outside wall of the AC and ns pulse excited coaxial DBD reaches 158 °C and 64.3 °C after 900 s operation, respectively. The experimental results on the comparison of the discharge characteristics of coaxial DBDs excited by different powers are significant for understanding of the mechanism of DBDs, reducing energy loss, and optimizing the performance of coaxial DBD in industrial applications.

  15. Photoabsorption cross sections of methane from 1400 to 1850 A

    NASA Technical Reports Server (NTRS)

    Mount, G. H.; Warden, E. S.; Moos, H. W.

    1977-01-01

    Photoabsorption cross sections of methane in the 1400-1850-A spectral region have been measured. Cross sections at wavelengths greater than 1475 A are approximately 200 times smaller than those currently accepted. This has a significant effect on the interpretation of spectral measurements of the Jovian planets in this wavelength region.

  16. Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

    NASA Astrophysics Data System (ADS)

    Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

    2017-10-01

    The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

  17. Investigation of materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  18. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  19. Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

    NASA Technical Reports Server (NTRS)

    Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

    1976-01-01

    Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

  20. [Evaluation of the capacity of the APR-DRG classification system to predict hospital mortality].

    PubMed

    De Marco, Maria Francesca; Lorenzoni, Luca; Addari, Piero; Nante, Nicola

    2002-01-01

    Inpatient mortality has increasingly been used as an hospital outcome measure. Comparing mortality rates across hospitals requires adjustment for patient risks before making inferences about quality of care based on patient outcomes. Therefore it is essential to dispose of well performing severity measures. The aim of this study is to evaluate the ability of the All Patient Refined DRG system to predict inpatient mortality for congestive heart failure, myocardial infarction, pneumonia and ischemic stroke. Administrative records were used in this analysis. We used two statistics methods to assess the ability of the APR-DRG to predict mortality: the area under the receiver operating characteristics curve (referred to as the c-statistic) and the Hosmer-Lemeshow test. The database for the study included 19,212 discharges for stroke, pneumonia, myocardial infarction and congestive heart failure from fifteen hospital participating in the Italian APR-DRG Project. A multivariate analysis was performed to predict mortality for each condition in study using age, sex and APR-DRG risk mortality subclass as independent variables. Inpatient mortality rate ranges from 9.7% (pneumonia) to 16.7% (stroke). Model discrimination, calculated using the c-statistic, was 0.91 for myocardial infarction, 0.68 for stroke, 0.78 for pneumonia and 0.71 for congestive heart failure. The model calibration assessed using the Hosmer-Leme-show test was quite good. The performance of the APR-DRG scheme when used on Italian hospital activity records is similar to that reported in literature and it seems to improve by adding age and sex to the model. The APR-DRG system does not completely capture the effects of these variables. In some cases, the better performance might be due to the inclusion of specific complications in the risk-of-mortality subclass assignment.

  1. 81. Neg. No. F63, Apr 13, 1930, INTERIORASSEMBLY BUILDING, HOOD ...

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

    81. Neg. No. F-63, Apr 13, 1930, INTERIOR-ASSEMBLY BUILDING, HOOD DEPARTMENT - Ford Motor Company Long Beach Assembly Plant, Assembly Building, 700 Henry Ford Avenue, Long Beach, Los Angeles County, CA

  2. A citation-based assessment of the performance of U.S. boiling water reactors following extended power up-rates

    NASA Astrophysics Data System (ADS)

    Heidrich, Brenden J.

    Nuclear power plants produce 20 percent of the electricity generated in the U.S. Nuclear generated electricity is increasingly valuable to a utility because it can be produced at a low marginal cost and it does not release any carbon dioxide. It can also be a hedge against uncertain fossil fuel prices. The construction of new nuclear power plants in the U.S. is cautiously moving forward, restrained by high capital costs. Since 1998, nuclear utilities have been increasing the power output of their reactors by implementing extended power up-rates. Power increases of up to 20 percent are allowed under this process. The equivalent of nine large power plants has been added via extended power up-rates. These up-rates require the replacement of large capital equipment and are often performed in concert with other plant life extension activities such as license renewals. This dissertation examines the effect of these extended power up-rates on the safety performance of U.S. boiling water reactors. Licensing event reports are submitted by the utilities to the Nuclear Regulatory Commission, the federal nuclear regulator, for a wide range of abnormal events. Two methods are used to examine the effect of extended power up-rates on the frequency of abnormal events at the reactors. The Crow/AMSAA model, a univariate technique is used to determine if the implementation of an extended power up-rate affects the rate of abnormal events. The method has a long history in the aerospace industry and in the military. At a 95-percent confidence level, the rate of events requiring the submission of a licensing event report decreases following the implementation of an extended power up-rate. It is hypothesized that the improvement in performance is tied to the equipment replacement and refurbishment that is performed as part of the up-rate process. The reactor performance is also analyzed using the proportional hazards model. This technique allows for the estimation of the effects of

  3. 47 CFR 95.1121 - Specific requirements for wireless medical telemetry devices operating in the 1395-1400 and 1427...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... wireless medical telemetry devices operating in the 1395-1400 and 1427-1432 MHz bands. Due to the critical... 47 Telecommunication 5 2010-10-01 2010-10-01 false Specific requirements for wireless medical telemetry devices operating in the 1395-1400 and 1427-1432 MHz bands. 95.1121 Section 95.1121...

  4. Heat pipe nuclear reactor for space power

    NASA Technical Reports Server (NTRS)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  5. Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

    NASA Astrophysics Data System (ADS)

    Ivanov, Yu. A.

    2007-12-01

    An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.

  6. 46 CFR 112.05-5 - Emergency power source.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... generator must be either a diesel engine or a gas turbine. [CGD 74-125A, 47 FR 15267, Apr. 8, 1982, as... power source (automatically connected storage battery or an automatically started generator) 36 hours.1... power source (automatically connected storage battery or an automatically started generator) 8 hours or...

  7. 46 CFR 112.05-5 - Emergency power source.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... generator must be either a diesel engine or a gas turbine. [CGD 74-125A, 47 FR 15267, Apr. 8, 1982, as... power source (automatically connected storage battery or an automatically started generator) 36 hours.1... power source (automatically connected storage battery or an automatically started generator) 8 hours or...

  8. The Alberta Pregnancy Outcomes and Nutrition (APrON) cohort study: rationale and methods.

    PubMed

    Kaplan, Bonnie J; Giesbrecht, Gerald F; Leung, Brenda M Y; Field, Catherine J; Dewey, Deborah; Bell, Rhonda C; Manca, Donna P; O'Beirne, Maeve; Johnston, David W; Pop, Victor J; Singhal, Nalini; Gagnon, Lisa; Bernier, Francois P; Eliasziw, Misha; McCargar, Linda J; Kooistra, Libbe; Farmer, Anna; Cantell, Marja; Goonewardene, Laki; Casey, Linda M; Letourneau, Nicole; Martin, Jonathan W

    2014-01-01

    The Alberta Pregnancy Outcomes and Nutrition (APrON) study is an ongoing prospective cohort study that recruits pregnant women early in pregnancy and, as of 2012, is following up their infants to 3 years of age. It has currently enrolled approximately 5000 Canadians (2000 pregnant women, their offspring and many of their partners). The primary aims of the APrON study were to determine the relationships between maternal nutrient intake and status, before, during and after gestation, and (1) maternal mood; (2) birth and obstetric outcomes; and (3) infant neurodevelopment. We have collected comprehensive maternal nutrition, anthropometric, biological and mental health data at multiple points in the pregnancy and the post-partum period, as well as obstetrical, birth, health and neurodevelopmental outcomes of these pregnancies. The study continues to follow the infants through to 36 months of age. The current report describes the study design and methods, and findings of some pilot work. The APrON study is a significant resource with opportunities for collaboration. © 2012 John Wiley & Sons Ltd.

  9. 83. Neg. No. F53, Apr 13, 1930, INTERIORASSEMBLY BUILDING, BACK ...

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

    83. Neg. No. F-53, Apr 13, 1930, INTERIOR-ASSEMBLY BUILDING, BACK TRIM LINE - Ford Motor Company Long Beach Assembly Plant, Assembly Building, 700 Henry Ford Avenue, Long Beach, Los Angeles County, CA

  10. 86. Neg. No. F64, Apr 13, 1930, INTERIORASSEMBLY BUILDING, BODY ...

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

    86. Neg. No. F-64, Apr 13, 1930, INTERIOR-ASSEMBLY BUILDING, BODY STORAGE CONVEYOR - Ford Motor Company Long Beach Assembly Plant, Assembly Building, 700 Henry Ford Avenue, Long Beach, Los Angeles County, CA

  11. Reactor monitoring using antineutrino detectors

    NASA Astrophysics Data System (ADS)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  12. 43. FLOOR PLAN OF POWER HOUSE, EXHIBIT L, SANTA ANA ...

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

    43. FLOOR PLAN OF POWER HOUSE, EXHIBIT L, SANTA ANA RIVER NO. 2 PROJECT, APR. 30, 1945. SCE drawing no. 523643 (sheet no. 14; for filing with Federal Power Commission). - Santa Ana River Hydroelectric System, SAR-2 Powerhouse, Redlands, San Bernardino County, CA

  13. 44. SECTIONS OF POWER HOUSE, EXHIBIT L, SANTA ANA RIVER ...

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

    44. SECTIONS OF POWER HOUSE, EXHIBIT L, SANTA ANA RIVER NO. 2 PROJECT, APR. 30, 1945. SCE drawing no. 523644 (sheet no. 15; for filing with Federal Power Commission). - Santa Ana River Hydroelectric System, SAR-2 Powerhouse, Redlands, San Bernardino County, CA

  14. 55. CROSS SECTION OF POWER HOUSE, EXHIBIT L, SANTA ANA ...

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

    55. CROSS SECTION OF POWER HOUSE, EXHIBIT L, SANTA ANA RIVER NO. 1 PROJECT, APR. 30, 1945. SCE drawing no. 523199 (sheet no. 9, for filing with Federal Power Commission). - Santa Ana River Hydroelectric System, SAR-1 Powerhouse, Redlands, San Bernardino County, CA

  15. Reliability and safety of the electrical power supply complex of the Hanford production reactors

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

    Robbins, F.D.

    Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high ordermore » of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.« less

  16. 15 CFR 782.1 - Overview of reporting requirements under the APR.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Foreign Trade (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL... (see § 782.5 of the APR). In addition, forms may be downloaded from the Internet at http://www.ap.gov. ...

  17. 15 CFR 782.1 - Overview of reporting requirements under the APR.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Foreign Trade (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL... (see § 782.5 of the APR). In addition, forms may be downloaded from the Internet at http://www.ap.gov. ...

  18. 15 CFR 782.1 - Overview of reporting requirements under the APR.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Foreign Trade (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL... (see § 782.5 of the APR). In addition, forms may be downloaded from the Internet at http://www.ap.gov. ...

  19. 15 CFR 782.1 - Overview of reporting requirements under the APR.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Foreign Trade (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL... (see § 782.5 of the APR). In addition, forms may be downloaded from the Internet at http://www.ap.gov. ...

  20. 85. Neg. No. F51, Apr 13, 1930, INTERIORASSEMBLY BUILDING, BODY ...

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

    85. Neg. No. F-51, Apr 13, 1930, INTERIOR-ASSEMBLY BUILDING, BODY AND CUSHION LINE - Ford Motor Company Long Beach Assembly Plant, Assembly Building, 700 Henry Ford Avenue, Long Beach, Los Angeles County, CA

  1. Assistant Personal Robot (APR): Conception and Application of a Tele-Operated Assisted Living Robot.

    PubMed

    Clotet, Eduard; Martínez, Dani; Moreno, Javier; Tresanchez, Marcel; Palacín, Jordi

    2016-04-28

    This paper presents the technical description, mechanical design, electronic components, software implementation and possible applications of a tele-operated mobile robot designed as an assisted living tool. This robotic concept has been named Assistant Personal Robot (or APR for short) and has been designed as a remotely telecontrolled robotic platform built to provide social and assistive services to elderly people and those with impaired mobility. The APR features a fast high-mobility motion system adapted for tele-operation in plain indoor areas, which incorporates a high-priority collision avoidance procedure. This paper presents the mechanical architecture, electrical fundaments and software implementation required in order to develop the main functionalities of an assistive robot. The APR uses a tablet in order to implement the basic peer-to-peer videoconference and tele-operation control combined with a tactile graphic user interface. The paper also presents the development of some applications proposed in the framework of an assisted living robot.

  2. Startup thaw concept for the SP-100 space reactor power system

    NASA Technical Reports Server (NTRS)

    Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

    1990-01-01

    A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

  3. 2 CFR 1400.970 - Nonprocurement transaction (Department of the Interior supplement to the definition at 2 CFR 180...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 2 Grants and Agreements 1 2011-01-01 2011-01-01 false Nonprocurement transaction (Department of the Interior supplement to the definition at 2 CFR 180.970). 1400.970 Section 1400.970 Grants and... supplement to the definition at 2 CFR 180.970). In addition to those listed in 2 CFR 180.970, the Department...

  4. 84. Neg. No. F62, Apr 13, 1930, INTERIORASSEMBLY BUILDING, FRAME ...

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

    84. Neg. No. F-62, Apr 13, 1930, INTERIOR-ASSEMBLY BUILDING, FRAME AND MOTOR STORAGE CONVEYOR - Ford Motor Company Long Beach Assembly Plant, Assembly Building, 700 Henry Ford Avenue, Long Beach, Los Angeles County, CA

  5. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  6. 46. LONGITUDINAL SECTION POWER HOUSE S.A.R. NO. 2, EDISON ...

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

    46. LONGITUDINAL SECTION - POWER HOUSE S.A.R. NO. 2, EDISON PROJECT, APR. 30, 1945. SCE drawing no. 523198 (sheet no. 8; for filing with Federal Power Commission). - Santa Ana River Hydroelectric System, SAR-2 Powerhouse, Redlands, San Bernardino County, CA

  7. Thermoelectric converter for SP-100 space reactor power system

    NASA Technical Reports Server (NTRS)

    Terrill, W. R.; Haley, V. F.

    1986-01-01

    Conductively coupling the thermoelectric converter to the heat source and the radiator maximizes the utilization of the reactor and radiator temperatures and thereby minimizes the power system weight. This paper presents the design for the converter and the individual thermoelectric cells that are the building block modules for the converter. It also summarizes progress on the fabrication of initial cells and the results obtained from the preparation of a manufacturing plan. The design developed for the SP-100 system utilizes thermally conductive compliant pads that can absorb the displacement and distortion caused by the combinations of temperatures and thermal expansion coefficients. The converter and cell designs provided a 100 kWe system which met the system requirements. Initial cells were fabricated and tested.

  8. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    DOEpatents

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  9. An experimental study of catalytic and non-catalytic reaction in heat recirculating reactors and applications to power generation

    NASA Astrophysics Data System (ADS)

    Ahn, Jeongmin

    An experimental study of the performance of a Swiss roll heat exchanger and reactor was conducted, with emphasis on the extinction limits and comparison of results with and without Pt catalyst. At Re<40, the catalyst was required to sustain reaction; with the catalyst self-sustaining reaction could be obtained at Re less than 1. Both lean and rich extinction limits were extended with the catalyst, though rich limits were extended much further. At low Re, the lean extinction limit was rich of stoichiometric and rich limit had equivalence ratios 80 in some cases. Non-catalytic reaction generally occurred in a flameless mode near the center of the reactor. With or without catalyst, for sufficiently robust conditions, a visible flame would propagate out of the center, but this flame could only be re-centered with catalyst. Gas chromatography indicated that at low Re, CO and non-C3 H8 hydrocarbons did not form. For higher Re, catalytic limits were slightly broader but had much lower limit temperatures. At sufficiently high Re, catalytic and gas-phase limits merged. Experiments with titanium Swiss rolls have demonstrated reducing wall thermal conductivity and thickness leads to lower heat losses and therefore increases operating temperatures and extends flammability limits. By use of Pt catalysts, reaction of propane-air mixtures at temperatures 54°C was sustained. Such low temperatures suggest that polymers may be employed as a reactor material. A polyimide reactor was built and survived prolonged testing at temperatures up to 500°C. Polymer reactors may prove more practical for microscale devices due to their lower thermal conductivity and ease of manufacturing. Since the ultimate goal of current efforts is to develop combustion driven power generation devices at MEMS like scales, a thermally self-sustaining miniature power generation device was developed utilizing a single-chamber solid-oxide-fuel-cell (SOFC) placed in a Swiss roll. With the single-chamber design

  10. 76. Neg. No. F58, Apr 13, 1930, INTERIORASSEMBLY BUILDING, BURNOFF, ...

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

    76. Neg. No. F-58, Apr 13, 1930, INTERIOR-ASSEMBLY BUILDING, BURNOFF, LOAD END OF ENAMEL OVEN - Ford Motor Company Long Beach Assembly Plant, Assembly Building, 700 Henry Ford Avenue, Long Beach, Los Angeles County, CA

  11. POWER BREEDER REACTOR

    DOEpatents

    Monson, H.O.

    1960-11-22

    An arrangement is offered for preventing or minimizing the contraction due to temperature rise, of a reactor core comprising vertical fuel rods in sodium. Temperature rise of the fuel rods would normally make them move closer together by inward bowing, with a resultant undesired increase in reactivity. According to the present invention, assemblies of the fuel rods are laterally restrained at the lower ends of their lower blanket sections and just above the middle of the fuel sections proper of the rods, and thus the fuel sections move apart, rather than together, with increase in temperature.

  12. New tools for NTD vaccines: A case study of quality control assays for product development of the human hookworm vaccine Na-APR-1M74.

    PubMed

    Pearson, Mark S; Jariwala, Amar R; Abbenante, Giovanni; Plieskatt, Jordan; Wilson, David; Bottazzi, Maria Elena; Hotez, Peter J; Keegan, Brian; Bethony, Jeffrey M; Loukas, Alex

    2015-01-01

    Na-APR-1(M74) is an aspartic protease that is rendered enzymatically inactive by site-directed mutagenesis and is a candidate antigen component in the Human Hookworm Vaccine. The mutant protease exerts vaccine efficacy by inducing antibodies that neutralize the enzymatic activity of wild type enzyme (Na-APR-1wt) in the gut of the hookworm, thereby depriving the worm of its ability to digest its blood meal. Previously, canines immunized with Na-APR-1(M74) and challenged with Ancylostoma caninum were partially protected against hookworm challenge infection, especially from the loss in hemoglobin observed in control canines and canine immunoglobulin (Ig) G raised against Na-APR-1 was shown to inhibit the enzymatic activity of Na-APR-1 wt in vitro, thereby providing proof of concept of Na-APR-1(M74) as a vaccine antigen. The mutated version, Na-APR-1(M74), was then expressed at the cGMP level using a Nicotiana benthamiana expression system (Fraunhofer, CMB, Delaware, MD), formulated with Alhydrogel®, and used to immunize mice in a dose-ranging study to explore the enzyme-neutralizing capacity of the resulting anti- Na-APR-1(M74) IgG. As little as 0.99 μg of recombinant Na-APR-1(M74) could induce anti Na-APR-1(M74) IgG in mice that were capable of inhibiting Na-APR-1w t-mediated digestion of a peptide substrate by 89%. In the absence of enzymatic activity of Na-APR-1(M74) as a surrogate marker of protein functionality, we developed an assay based on the binding of a quenched fluorescence-labeled inhibitor of aspartic proteases, BODIPY-FL pepstatin A (BDP). Binding of BDP in the active site of Na-APR-1 wt was demonstrated by inhibition of enzymatic activity, and competitive binding with unlabelled pepstatin A. BDP also bound to Na-APR-1(M74) which was assessed by fluorescence polarization, but with an ∼ 50-fold reduction in the dissociation constant. Taken together, these assays comprise a "toolbox" that could be useful for the analyses of Na-APR-1(M74) as it

  13. New tools for NTD vaccines: A case study of quality control assays for product development of the human hookworm vaccine Na-APR-1M74

    PubMed Central

    Pearson, Mark S; Jariwala, Amar R; Abbenante, Giovanni; Plieskatt, Jordan; Wilson, David; Bottazzi, Maria Elena; Hotez, Peter J; Keegan, Brian; Bethony, Jeffrey M; Loukas, Alex

    2015-01-01

    Na-APR-1M74 is an aspartic protease that is rendered enzymatically inactive by site-directed mutagenesis and is a candidate antigen component in the Human Hookworm Vaccine. The mutant protease exerts vaccine efficacy by inducing antibodies that neutralize the enzymatic activity of wild type enzyme (Na-APR-1wt) in the gut of the hookworm, thereby depriving the worm of its ability to digest its blood meal. Previously, canines immunized with Na-APR-1M74 and challenged with Ancylostoma caninum were partially protected against hookworm challenge infection, especially from the loss in hemoglobin observed in control canines and canine immunoglobulin (Ig) G raised against Na-APR-1 was shown to inhibit the enzymatic activity of Na-APR-1wt in vitro, thereby providing proof of concept of Na-APR-1M74 as a vaccine antigen. The mutated version, Na-APR-1M74, was then expressed at the cGMP level using a Nicotiana benthamiana expression system (Fraunhofer, CMB, Delaware, MD), formulated with Alhydrogel®, and used to immunize mice in a dose-ranging study to explore the enzyme-neutralizing capacity of the resulting anti- Na-APR-1M74 IgG. As little as 0.99 μg of recombinant Na-APR-1M74 could induce anti Na-APR-1M74 IgG in mice that were capable of inhibiting Na-APR-1wt-mediated digestion of a peptide substrate by 89%. In the absence of enzymatic activity of Na-APR-1M74 as a surrogate marker of protein functionality, we developed an assay based on the binding of a quenched fluorescence-labeled inhibitor of aspartic proteases, BODIPY-FL pepstatin A (BDP). Binding of BDP in the active site of Na-APR-1wt was demonstrated by inhibition of enzymatic activity, and competitive binding with unlabelled pepstatin A. BDP also bound to Na-APR-1M74 which was assessed by fluorescence polarization, but with an ∼50-fold reduction in the dissociation constant. Taken together, these assays comprise a “toolbox” that could be useful for the analyses of Na-APR-1M74 as it proceeds through the

  14. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

    DOEpatents

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

    2014-11-18

    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  15. Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

    NASA Astrophysics Data System (ADS)

    Tyapkov, V. F.; Erpyleva, S. F.

    2017-05-01

    Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

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

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

    Hursin, M.; Perret, G.

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

  17. 15 CFR 782.5 - Where to obtain APR report forms.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL REGULATIONS GENERAL... forms required by the APR may be downloaded from the Internet at http://www.ap.gov. You also may obtain these forms by contacting: Treaty Compliance Division, Bureau of Industry and Security, U.S. Department...

  18. 15 CFR 782.5 - Where to obtain APR report forms.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL REGULATIONS GENERAL... forms required by the APR may be downloaded from the Internet at http://www.ap.gov. You also may obtain these forms by contacting: Treaty Compliance Division, Bureau of Industry and Security, U.S. Department...

  19. 15 CFR 782.5 - Where to obtain APR report forms.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL REGULATIONS GENERAL... forms required by the APR may be downloaded from the Internet at http://www.ap.gov. You also may obtain these forms by contacting: Treaty Compliance Division, Bureau of Industry and Security, U.S. Department...

  20. 15 CFR 782.5 - Where to obtain APR report forms.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... (Continued) BUREAU OF INDUSTRY AND SECURITY, DEPARTMENT OF COMMERCE ADDITIONAL PROTOCOL REGULATIONS GENERAL... forms required by the APR may be downloaded from the Internet at http://www.ap.gov. You also may obtain these forms by contacting: Treaty Compliance Division, Bureau of Industry and Security, U.S. Department...