Sample records for reactors protection systems

  1. A novel plant protection strategy for transient reactors

    NASA Astrophysics Data System (ADS)

    Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

    A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

  2. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    DTIC Science & Technology

    2011-03-01

    protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of...conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to...protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader

  3. A novel plant protection strategy for transient reactors

    NASA Astrophysics Data System (ADS)

    Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

    The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

  4. Integral isolation valve systems for loss of coolant accident protection

    DOEpatents

    Kanuch, David J.; DiFilipo, Paul P.

    2018-03-20

    A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

  5. 10 CFR 50.48 - Fire protection.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... suppression systems; and (iii) The means to limit fire damage to structures, systems, or components important...) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating... pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis...

  6. 10 CFR 50.48 - Fire protection.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... suppression systems; and (iii) The means to limit fire damage to structures, systems, or components important...) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating... pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis...

  7. TREAT Reactor Control and Protection System

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

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

  8. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, Donald C.

    1996-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

  9. Reactor protection system with automatic self-testing and diagnostic

    DOEpatents

    Gaubatz, D.C.

    1996-12-17

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

  10. 77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-11

    ...-492- 3668; email: [email protected] . NRC's Agencywide Documents Access and Management System... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand...

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

    Soubies, B.; Henry, J.Y.; Le Meur, M.

    1300 MWe pressurised water reactors (PWRs), like the 1400 MWe reactors, operate with microprocessor-based safety systems. This is particularly the case for the Digital Integrated Protection System (SPIN), which trips the reactor in an emergency and sets in action the safeguard functions. The softwares used in these systems must therefore be highly dependable in the execution of their functions. In the case of SPIN, three players are working at different levels to achieve this goal: the protection system manufacturer, Merlin Gerin; the designer of the nuclear steam supply system, Framatome; the operator of the nuclear power plants, Electricite de Francemore » (EDF), which is also responsible for the safety of its installations. Regulatory licenses are issued by the French safety authority, the Nuclear Installations Safety Directorate (French abbreviation DSIN), subsequent to a successful examination of the technical provisions adopted by the operator. This examination is carried out by the IPSN and the standing group on nuclear reactors. This communication sets out: the methods used by the manufacturer to develop SPIN software for the 1400 MWe PWRs (N4 series); the approach adopted by the IPSN to evaluate the safety software of the protection system for the N4 series of reactors.« less

  12. 76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-08-08

    ... nuclear reactor facility. PBAPS Unit 1 was a high-temperature, gas-cooled reactor that was operated from... the safeguards contingency plan.'' Part 73 of 10 CFR, ``Physical Protection of Plant and Materials... physical protection system which will have capabilities for the protection of special nuclear material at...

  13. Autonomous Control of Space Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

  14. 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.

  15. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  16. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  17. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, Donald C.

    1997-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

  18. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, D.C.

    1997-04-15

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

  19. Army gas-cooled reactor systems program. Preliminary design report off-normal scram system

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

    Bushnell, W.H.; Malmstrom, S.A.

    1965-06-01

    The maximum allowable ML-1 fuel element cladding (hot spot) temperature is established by ANTS 201 at 1750/sup 0/F. The existing ML-1 design makes no provision for automatic scram when this limit is reached. Operating experience has indicated a requirement for such an automatic system during plant startup and a revised hot spot envelope (generated during conceptual design of the scram system) established the desirability of extending this protection to operation at full power conditions. It was also determined that the scram system should include circuitry to initiate an automatic scram if reactor ..delta..T exceeded 450/sup 0/F (the limit established inmore » ANTS 201) and if reactor power exceeded 6 kw(t) without coolant flow in the main loop. The preliminary design of the scram system (designated off-normal scram system) which will provide the required protection is described.« less

  20. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

    1989-01-01

    A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

  1. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  2. Thermal margin protection system for a nuclear reactor

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

    Musick, C.R.

    1974-02-12

    A thermal margin protection system for a nuclear reactor is described where the coolant flow flow trip point and the calculated thermal margin trip point are switched simultaneously and the thermal limit locus is made more restrictive as the allowable flow rate is decreased. The invention is characterized by calculation of the thermal limit Locus in response to applied signals which accurately represent reactor cold leg temperature and core power; cold leg temperature being corrected for stratification before being utilized and reactor power signals commensurate with power as a function of measured neutron flux and thermal energy added to themore » coolant being auctioneered to select the more conservative measure of power. The invention further comprises the compensation of the selected core power signal for the effects of core radial peaking factor under maximum coolant flow conditions. (Official Oazette)« less

  3. 10 CFR 140.12 - Amount of financial protection required for other reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

  4. 10 CFR 140.12 - Amount of financial protection required for other reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

  5. 10 CFR 140.12 - Amount of financial protection required for other reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

  6. 10 CFR 140.12 - Amount of financial protection required for other reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

  7. 10 CFR 140.12 - Amount of financial protection required for other reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

  8. Design and testing of a self-actuated shut down system for the protection of liquid metal fast breeder reactors (LMFBRs)

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

    Josephson, J.; Sowa, E.S.

    1977-04-01

    The design and testing of a simple and reliable Self-Actuated Shutdown System (SASS) for the protection of Liquid Metal Fast Breeder Reactors (LMFBRs) is described. A ferromagnetic Curie temperature permanent magnet holding device has been selected for the design of the Self-Actuated Shutdown System in order to enhance the safety of liquid metal cooled fast reactors (LMFBRs). The self-actuated, self-contained device operates such that accident conditions, resulting in increased coolant temperature or neutron flux reduce the magnetic holding force suspending a neutron absorber above the core by raising the temperature of the trigger mechanism above the Curie point. Neutron absorbermore » material is then inserted into the core, under gravity, terminating the accident. Two possible design variations of the selected concept are presented.« less

  9. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  10. Integrated intelligent systems in advanced reactor control rooms

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

    Beckmeyer, R.R.

    1989-01-01

    An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs.,more » 5 figs.« less

  11. Class notes from the first international training course on the physical protection of nuclear facilities and materials

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

    Herrington, P.B.

    1979-05-01

    The International Training Course on Physical Protection of Nuclear Facilities and Materials was intended for representatives from the developing countries who are responsible for preparing regulations and designing and assessing physical protection systems. The first part of the course consists of lectures on the objectives, organizational characteristics, and licensing and regulations requirements of a state system of physical protection. Since the participants may have little experience in nuclear energy, background information is provided on the topics of nuclear materials, radiation hazards, reactor systems, and reactor operations. Transportation of nuclear materials is addressed and emphasis is placed on regulations. Included inmore » these discussions are presentations by guest speakers from countries outside the United States of America who present their countries' threat to nuclear facilities. Effectiveness evaluation methodology is introduced to the participants by means of instructions which teach them how to use logic trees and the EASI (Estimate of Adversary Sequence Interruption) program. The following elements of a physical protection system are discussed: barriers, protective force, intrusion detection systems, communications, and entry-control systems. Total systems concepts of physical protection system design are emphasized throughout the course. Costs, manpower/technology trade-offs, and other practical considerations are discussed. Approximately one-third of the course is devoted to practical exercises during which the attendees participatein problem solving. A hypothetical nuclear facility is introduced, and the attendees participate in the conceptual design of a physical protection system for the facility.« less

  12. On the possibility of connecting a non-operating main circulation pump with three pumps in operation without preliminary coast-down of power-generating unit No. 5 in the Novovoronezh nuclear power plant

    NASA Astrophysics Data System (ADS)

    Vitkovskii, I. L.; Nikonov, S. P.; Ryasnyi, S. I.

    2014-02-01

    The subject of this paper is a transient caused by connection of a standby loop to three operating circulation pumps at the initial reactor heat rate equal to 70% of the rated value without preliminarily reducing it to 30% of the rated level as required by the safe operation regulations. Failure of the following normal operation systems is supposed: the first- and the second-type warning protection systems, all quick-acting reducing devices releasing steam into the auxiliary manifold, the electric heaters of the pressurizer, the pressurizer injection system, the primary cooling circuit fluid makeup/blow-through systems, and the blocking systems to shut down the main circulation pump after the level in the steam generator is exceeded. In addition, it is supposed that, under transient conditions, the valves of the turbine regulation system will be in the position in which they were at the moment of the initial event until generation of the signal for positive closing of the turbine stop valves. The first signal to actuate the reactor emergency protection system (EPS) is skipped. The failure of all quick-acting reducing devices releasing steam into the atmosphere is assumed. In addition to equipment failure, at the moment when the main circulation pump is connected, the operator erroneously puts in a new setting to maintain the power allowable for four pumps in operation-in the calculations it was taken equal to 104% of the rated level at most considering the accuracy of evaluating and maintaining the reactor heat rate-and the working group of the reactor protection and control system (P&CS) starts moving upward. On reaching the set power level, the automatic reactor power regulator stops operating and the P&CS elements remain in the position in which they are at the moment. Compliance with the design safety criteria for the adopted scenario of the transient is demonstrated.

  13. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

  14. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

  15. Automatic reactor control system for transient operation

    NASA Astrophysics Data System (ADS)

    Lipinski, Walter C.; Bhattacharyya, Samit K.; Hanan, Nelson A.

    Various programmatic considerations have delayed the upgrading of the TREAT reactor and the performance of the control system is not yet experimentally verified. The current schedule calls for the upgrading activities to occur last in the calendar year 1987. Detailed simulation results, coupled with earlier validation of individual components of the control strategy in TREAT, verify the performance of the algorithms. The control system operates within the safety envelope provided by a protection system designed to ensure reactor safety under conditions of spurious reactivity additions. The approach should be directly applicable to MMW systems, with appropriate accounting of temperature rate limitations of key components and of the inertia of the secondary system components.

  16. Dense, layered membranes for hydrogen separation

    DOEpatents

    Roark, Shane E.; MacKay, Richard; Mundschau, Michael V.

    2006-02-21

    This invention provides hydrogen-permeable membranes for separation of hydrogen from hydrogen-containing gases. The membranes are multi-layer having a central hydrogen-permeable layer with one or more catalyst layers, barrier layers, and/or protective layers. The invention also relates to membrane reactors employing the hydrogen-permeable membranes of the invention and to methods for separation of hydrogen from a hydrogen-containing gas using the membranes and reactors. The reactors of this invention can be combined with additional reactor systems for direct use of the separated hydrogen.

  17. Lunar Surface Reactor Shielding Study

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

    Kang, Shawn; McAlpine, William; Lipinski, Ronald

    A nuclear reactor system could provide power to support long term human exploration of the moon. Such a system would require shielding to protect astronauts from its emitted radiations. Shielding studies have been performed for a Gas Cooled Reactor system because it is considered to be the most suitable nuclear reactor system available for lunar exploration, based on its tolerance of oxidizing lunar regolith and its good conversion efficiency. The goals of the shielding studies were to determine a material shielding configuration that reduces the dose (rem) to the required level in order to protect astronauts, and to estimate themore » mass of regolith that would provide an equivalent protective effect if it were used as the shielding material. All calculations were performed using MCNPX, a Monte Carlo transport code. Lithium hydride must be kept between 600 K and 700 K to prevent excessive swelling from large amounts of gamma or neutron irradiation. The issue is that radiation damage causes separation of the lithium and the hydrogen, resulting in lithium metal and hydrogen gas. The proposed design uses a layer of B4C to reduce the combined neutron and gamma dose to below 0.5Grads before the LiH is introduced. Below 0.5Grads the swelling in LiH is small (less than about 1%) for all temperatures. This approach causes the shield to be heavier than if the B4C were replaced by LiH, but it makes the shield much more robust and reliable.« less

  18. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... integration of systems, technologies, programs, equipment, supporting processes, and implementing procedures...-in-depth methodologies to minimize the potential for an insider to adversely affect, either directly... protection of digital computer and communication systems and networks. (ii) Site-specific conditions that...

  19. AQUABOX 50 AND MARABU PACKED BIOLOGICAL REACTOR SYSTEM TECHNOLOGY EVALUATION, STADTWERKE DUESSELDORF AG SITE, DUESSELDORF, GERMANY

    EPA Science Inventory

    This ITER summarizes the results of an evaluation of the AQUABOX 50 and MARABU Packed Biological Reactor technologies. The evaluation was conducted under a bilateral agreement between the United States (U.S.) Environmental Protection Agency (EPA) Superfund Innovative Technology ...

  20. Protection of biomass from snail overgrazing in a trickling filter using sponge media as a biomass carrier: down-flow hanging sponge system.

    PubMed

    Onodera, Takashi; Syutsubo, Kazuaki; Yoochatchaval, Wilasinee; Sumino, Haruhiko; Mizuochi, Motoyuki; Harada, Hideki

    2015-01-01

    This study investigated down-flow hanging sponge (DHS) technology as a promising trickling filter (TF) using sponge media as a biomass carrier with an emphasis on protection of the biomass against macrofauna overgrazing. A pilot-scale DHS reactor fed with low-strength municipal sewage was operated under ambient temperature conditions for 1 year at a sewage treatment plant in Bangkok, Thailand. The results showed that snails (macrofauna) were present on the surface of the sponge media, but could not enter into it, because the sponge media with smaller pores physically protected the biomass from the snails. As a result, the sponge media maintained a dense biomass, with an average value of 22.3 gVSS/L sponge (58.1 gTSS/L sponge) on day 370. The snails could graze biomass on the surface of the sponge media. The DHS reactor process performance was also successful. The DHS reactor requires neither chemical treatments nor specific operations such as flooding for snail control. Overall, the results of this study indicate that the DHS reactor is able to protect biomass from snail overgrazing.

  1. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

  2. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

  3. Advanced shield development for a fission surface power system for the lunar surface

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

    A. E. Craft; I. J. Silver; C. M. Clark

    A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermetmore » fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.« less

  4. 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

  5. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Amounts of financial protection for certain reactors. 140... reactors. (a) Each licensee is required to have and maintain financial protection: (1) In the amount of $1,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding...

  6. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Amounts of financial protection for certain reactors. 140... reactors. (a) Each licensee is required to have and maintain financial protection: (1) In the amount of $1,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding...

  7. Low-temperature overpressurization protection system setpoint analysis using RETRAN-02/MOD5 for Salem

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

    Dodson, R.J.; Feltus, M.A.

    The low-temperature overpressurization protection system (LTOPS) is designed to protect the reactor pressure vessel (RPV) from brittle failure during startup and cooldown maneuvers in Westinghouse pressurized water reactors. For the Salem power plants, the power-operated relief valves (PORVs) mitigate pressure increases above a setpoint where an operational startup transient may put the RPV in the embrittlement fracture zone. The Title 10, Part 50, Code of Federal Regulations Appendix G limit, given by plant technical specifications, conservatively bounds the maximum pressure allowed during those transients where the RPV can suffer brittle fracture (usually below 350{degrees}F). The Appendix G limit is amore » pressure versus temperature curve that is more restrictive at lower RPV temperatures and allows for higher pressures as the temperature approaches the upper bounding fracture temperature.« less

  8. Reactor safeguards system assessment and design. Volume I

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

    Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.

    1978-06-01

    This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore,more » should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems.« less

  9. SP-100 reactor with Brayton conversion for lunar surface applications

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Rodriguez, Carlos D.; Mckissock, Barbara I.; Hanlon, James C.; Mansfield, Brian C.

    1992-01-01

    Examined here is the potential for integrating Brayton-cycle power conversion with the SP-100 reactor for lunar surface power system applications. Two designs were characterized and modeled. The first design integrates a 100-kWe SP-100 Brayton power system with a lunar lander. This system is intended to meet early lunar mission power needs while minimizing on-site installation requirements. Man-rated radiation protection is provided by an integral multilayer, cylindrical lithium hydride/tungsten (LiH/W) shield encircling the reactor vessel. Design emphasis is on ease of deployment, safety, and reliability, while utilizing relatively near-term technology. The second design combines Brayton conversion with the SP-100 reactor in a erectable 550-kWe powerplant concept intended to satisfy later-phase lunar base power requirements. This system capitalizes on experience gained from operating the initial 100-kWe module and incorporates some technology improvements. For this system, the reactor is emplaced in a lunar regolith excavation to provide man-rated shielding, and the Brayton engines and radiators are mounted on the lunar surface and extend radially from the central reactor. Design emphasis is on performance, safety, long life, and operational flexibility.

  10. Planetary surface reactor shielding using indigenous materials

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

    Houts, Michael G.; Poston, David I.; Trellue, Holly R.

    The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials.

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

    Curry, J J; Gallagher, D W; Modarres, M

    Appendices are presented concerning isolation condenser makeup; vapor suppression system; station air system; reactor building closed cooling water system; turbine building secondary closed water system; service water system; emergency service water system; fire protection system; emergency ac power; dc power system; event probability estimation; methodology of accident sequence quantification; and assignment of dominant sequences to release categories.

  12. PBF Reactor Building (PER620) basement. Workers wearing protective gear work ...

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

    PBF Reactor Building (PER-620) basement. Workers wearing protective gear work inside cubicle 13 on the fission product detection system. Man on left is atop shielded box shown in previous photo. Posture of second man illustrates waist-high height of shielding box. His hand rests on the access panel, which has been filled with lead bricks and which has been slid shut to enclose detection instruments within box. Photographer: John Capek. Date: January 24, 1983. INEEL negative no. 83-41-3-5 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  13. 10 CFR 73.21 - Protection of Safeguards Information: Performance requirements.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... maintain an information protection system that includes the applicable measures for Safeguards Information specified in § 73.22 related to: Power reactors; a formula quantity of strategic special nuclear material; transportation of or delivery to a carrier for transportation of a formula quantity of strategic special nuclear...

  14. Institute of Electrical and Electronics Engineers, Nuclear Science Symposium, 18th, and Nuclear Power Systems Symposium, 3rd, San Francisco, Calif., November 3-5, 1971, Proceedings.

    NASA Technical Reports Server (NTRS)

    1972-01-01

    Potential advantages of fusion power reactors are discussed together with the protection of the public from radioactivity produced in nuclear power reactors, and the significance of tritium releases to the environment. Other subjects considered are biomedical instrumentation, radiation damage problems, low level environmental radionuclide analysis systems, nuclear techniques in environmental research, nuclear instrumentation, and space and plasma instrumentation. Individual items are abstracted in this issue.

  15. Fault tree applications within the safety program of Idaho Nuclear Corporation

    NASA Technical Reports Server (NTRS)

    Vesely, W. E.

    1971-01-01

    Computerized fault tree analyses are used to obtain both qualitative and quantitative information about the safety and reliability of an electrical control system that shuts the reactor down when certain safety criteria are exceeded, in the design of a nuclear plant protection system, and in an investigation of a backup emergency system for reactor shutdown. The fault tree yields the modes by which the system failure or accident will occur, the most critical failure or accident causing areas, detailed failure probabilities, and the response of safety or reliability to design modifications and maintenance schemes.

  16. Planetary surface reactor shielding using indigenous materials

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

    Houts, Michael G.; Poston, David I.; Trellue, Holly R.

    The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials. {copyright} {ital 1999 American Institute of Physics.}

  17. IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ...

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

    IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ion chamber beneath reactor on dolly. Photographer: Page Comiskey. Date: August 11, 1965. INEEL negative no. 65-4039 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  18. Preliminary consideration of CFETR ITER-like case diagnostic system.

    PubMed

    Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

    2016-11-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  19. Preliminary consideration of CFETR ITER-like case diagnostic system

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

    Li, G. S.; Liu, Y. K.; Gao, X.

    2016-11-15

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basicmore » control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.« less

  20. Reliability of digital reactor protection system based on extenics.

    PubMed

    Zhao, Jing; He, Ya-Nan; Gu, Peng-Fei; Chen, Wei-Hua; Gao, Feng

    2016-01-01

    After the Fukushima nuclear accident, safety of nuclear power plants (NPPs) is widespread concerned. The reliability of reactor protection system (RPS) is directly related to the safety of NPPs, however, it is difficult to accurately evaluate the reliability of digital RPS. The method is based on estimating probability has some uncertainties, which can not reflect the reliability status of RPS dynamically and support the maintenance and troubleshooting. In this paper, the reliability quantitative analysis method based on extenics is proposed for the digital RPS (safety-critical), by which the relationship between the reliability and response time of RPS is constructed. The reliability of the RPS for CPR1000 NPP is modeled and analyzed by the proposed method as an example. The results show that the proposed method is capable to estimate the RPS reliability effectively and provide support to maintenance and troubleshooting of digital RPS system.

  1. Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

    NASA Astrophysics Data System (ADS)

    Rebak, Raul B.

    2018-02-01

    The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

  2. Self-actuated shutdown system for a commercial size LMFBR. Final report

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

    Dupen, C.F.G.

    1978-08-01

    A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

  3. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

  4. 10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

  5. Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

    NASA Astrophysics Data System (ADS)

    Blanchet, David; Fontaine, Bruno

    2017-09-01

    The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

  6. Fukushima Daiichi Information Repository FY13 Status

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

    Smith, Curtis; Phelan, Cherie; Schwieder, Dave

    The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in commercial nuclear power plant operating history. Much will be learned that may be applicable to the U.S. reactor fleet, nuclear fuel cycle facilities, and supporting systems, and the international reactor fleet. For example, lessons from Fukushima Daiichi may be applied to emergency response planning, reactor operator training, accident scenario modeling, human factors engineering, radiation protection, and accident mitigation; as well as influence U.S. policies towards the nuclear fuel cycle including power generation, and spent fuel storage, reprocessing, and disposal. This document describesmore » the database used to establish a centralized information repository to store and manage the Fukushima data that has been gathered. The data is stored in a secured (password protected and encrypted) repository that is searchable and available to researchers at diverse locations.« less

  7. Method of shielding a liquid-metal-cooled reactor

    DOEpatents

    Sayre, Robert K.

    1978-01-01

    The primary heat transport system of a nuclear reactor -- particularly for a liquid-metal-cooled fast-breeder reactor -- is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system.

  8. PBF Reactor Building (PER620). Camera on main floor faces south ...

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

    PBF Reactor Building (PER-620). Camera on main floor faces south (open) doorway. In foreground is canal gate, lined with stainless steel and painted with protective coatings. Reactor pit is round with protective coatings. Reactor put is round form discernible beyond. Lifting beams and rigging are in place for a load test before reactor vessel arrives. Photographer: John Capek. Date: January 26, 1970. INEEL negative no. 70-347 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  9. The effectiveness of using the calculated braking current for longitudinal differential protection of 110 - 750 kV shunt reactors

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

    Vdovin, S. A.; Shalimov, A. S.

    2013-05-15

    The use of the function of effective current braking of the longitudinal differential protection of shunt reactors to offset current surges, which enables the sensitivity of differential protection to be increased when there are short circuits with low damage currents, is considered. It is shown that the use of the calculated braking characteristic enables the reliability of offset protection from transients to be increased when the reactor is connected, which is accompanied by the flow of asymmetric currents containing an aperiodic component.

  10. 40 CFR 63.11499 - What are the standards and compliance requirements for heat exchange systems?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... requirements for heat exchange systems? 63.11499 Section 63.11499 Protection of Environment ENVIRONMENTAL... and compliance requirements for heat exchange systems? (a) If the cooling water flow rate in your heat... part 211, you may use the physical integrity of the reactor as the surrogate indicator of heat...

  11. 40 CFR 63.11499 - What are the standards and compliance requirements for heat exchange systems?

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... requirements for heat exchange systems? 63.11499 Section 63.11499 Protection of Environment ENVIRONMENTAL... and compliance requirements for heat exchange systems? (a) If the cooling water flow rate in your heat... part 211, you may use the physical integrity of the reactor as the surrogate indicator of heat...

  12. 40 CFR 63.11499 - What are the standards and compliance requirements for heat exchange systems?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... requirements for heat exchange systems? 63.11499 Section 63.11499 Protection of Environment ENVIRONMENTAL... and compliance requirements for heat exchange systems? (a) If the cooling water flow rate in your heat... part 211, you may use the physical integrity of the reactor as the surrogate indicator of heat...

  13. 40 CFR 63.11499 - What are the standards and compliance requirements for heat exchange systems?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... requirements for heat exchange systems? 63.11499 Section 63.11499 Protection of Environment ENVIRONMENTAL... and compliance requirements for heat exchange systems? (a) If the cooling water flow rate in your heat... part 211, you may use the physical integrity of the reactor as the surrogate indicator of heat...

  14. 40 CFR 63.11499 - What are the standards and compliance requirements for heat exchange systems?

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... requirements for heat exchange systems? 63.11499 Section 63.11499 Protection of Environment ENVIRONMENTAL... and compliance requirements for heat exchange systems? (a) If the cooling water flow rate in your heat... part 211, you may use the physical integrity of the reactor as the surrogate indicator of heat...

  15. 76 FR 5102 - Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated Reactor Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... 3150-AI64 [NRC-2010-0340] Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated...-0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance on implementing the provisions of proposed 10 CFR Part 73.37, ``Requirements for Physical Protection...

  16. First flush reactor for stormwater treatment for elevated linear transportation projects.

    DOT National Transportation Integrated Search

    2009-06-01

    The United States EPA (Environmental Protection Agency) MS4 (Municipal Separate Storm Water Sewer System) Program regulations : require municipalities and government agencies including the Louisiana Department of Transportation and Development (LADOT...

  17. Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

    NASA Astrophysics Data System (ADS)

    Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

    2018-02-01

    The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

  18. Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

    NASA Astrophysics Data System (ADS)

    Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

    2017-12-01

    The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

  19. 75 FR 76923 - Domestic Licensing of Production and Utilization Facilities; Updates to Incorporation by...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-10

    ..., Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping...

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

    NASA Astrophysics Data System (ADS)

    McCormick, Kathy

    2013-10-01

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

  1. Surface Catalysis and Characterization of Proposed Candidate TPS for Access-to-Space Vehicles

    NASA Technical Reports Server (NTRS)

    Stewart, David A.

    1997-01-01

    Surface properties have been obtained on several classes of thermal protection systems (TPS) using data from both side-arm-reactor and arc-jet facilities. Thermochemical stability, optical properties, and coefficients for atom recombination were determined for candidate TPS proposed for single-stage-to-orbit vehicles. The systems included rigid fibrous insulations, blankets, reinforced carbon carbon, and metals. Test techniques, theories used to define arc-jet and side-arm-reactor flow, and material surface properties are described. Total hemispherical emittance and atom recombination coefficients for each candidate TPS are summarized in the form of polynomial and Arrhenius expressions.

  2. Developments and Tendencies in Fission Reactor Concepts

    NASA Astrophysics Data System (ADS)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

  3. Taking Steps to Protect Against the Insider Threat

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

    Pope, Noah Gale; Williams, Martha; Lewis, Joel

    2015-10-16

    Research reactors are required (in accordance with the Safeguards Agreement between the State and the IAEA) to maintain a system of nuclear material accounting and control for reporting quantities of nuclear material received, shipped, and held on inventory. Enhancements to the existing accounting and control system can be made at little additional cost to the facility, and these enhancements can make nuclear material accounting and control useful for nuclear security. In particular, nuclear material accounting and control measures can be useful in protecting against an insider who is intent on unauthorized removal or misuse of nuclear material or misuse ofmore » equipment. An enhanced nuclear material accounting and control system that responds to nuclear security is described in NSS-25G, Use of Nuclear Material Accounting and Control for Nuclear Security Purposes at Facilities, which is scheduled for distribution by the IAEA Department of Nuclear Security later this year. Accounting and control measures that respond to the insider threat are also described in NSS-33, Establishing a System for Control of Nuclear Material for Nuclear Security Purposes at a Facility During Storage, Use and Movement, and in NSS-41, Preventive and Protective Measures against Insider Threats (originally issued as NSS-08), which are available in draft form. This paper describes enhancements to existing material control and accounting systems that are specific to research reactors, and shows how they are important to nuclear security and protecting against an insider.« less

  4. Initial Assessment of Environmental Barrier Coatings for the Prometheus Project

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

    M. Frederick

    2005-12-15

    Depending upon final design and materials selections, a variety of engineering solutions may need to be considered to avoid chemical degradation of components in a notional space nuclear power plant (SNPP). Coatings are one engineered approach that was considered. A comprehensive review of protective coating technology for various space-reactor structural materials is presented, including refractory metal alloys [molybdenum (Mo), tungsten (W), rhenium (Re), tantalum (Ta), and niobium (Nb)], nickel (Ni)-base superalloys, and silicon carbide (Sic). A summary description of some common deposition techniques is included. A literature survey identified coatings based on silicides or iridium/rhenium as the primary methods formore » environmental protection of refractory metal alloys. Modified aluminide coatings have been identified for superalloys and multilayer ceramic coatings for protection of Sic. All reviewed research focused on protecting structural materials from extreme temperatures in highly oxidizing conditions. Thermodynamic analyses indicate that some of these coatings may not be protective in the high-temperature, impure-He environment expected in a Prometheus reactor system. Further research is proposed to determine extensibility of these coating materials to less-oxidizing or neutral environments.« less

  5. 75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-03

    ...-2010-0340; Draft NUREG-0561, Revision 2] RIN 3150-AI64 Physical Protection of Shipments of Irradiated...- 0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance to a licensee or applicant for implementation of proposed 10 CFR 73.37, ``Requirements for Physical...

  6. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-28

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 [NRC-2010-0340; NRC-2009-0163] RIN 3150-AI64 Physical..., ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This revised document sets forth means... physical protection of spent nuclear fuel (SNF) during transportation by road, rail, and water; and for...

  7. First flush reactor for stormwater treatment for elevated linear transportation projects : tech summary.

    DOT National Transportation Integrated Search

    2009-06-01

    The U.S. Environmental Protection Agency (EPA) Municipal Separate Storm Water Sewer System (MS4) Program : regulations require municipalities and government agencies including the Louisiana Department of Transportation : and Development (LADOTD) to d...

  8. 10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... fuel (100 grams or less) in transit. 73.35 Section 73.35 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of irradiated...

  9. Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

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

    Murav’ev, V. P., E-mail: murval1@mail.ru

    The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

  10. REACTOR VIEWING APPARATUS

    DOEpatents

    Monk, G.S.

    1959-01-13

    An optical system is presented that is suitable for viewing objects in a region of relatively high radioactivity, or high neutron activity, such as a neutronic reactor. This optical system will absorb neutrons and gamma rays thereby protecting personnel fronm the harmful biological effects of such penetrating radiations. The optical system is comprised of a viewing tube having a lens at one end, a transparent solid member at the other end and a transparent aqueous liquid completely filling the tube between the ends. The lens is made of a polymerized organic material and the transparent solid member is made of a radiation absorbent material. A shield surrounds the tube betwcen the flanges and is made of a gamma ray absorbing material.

  11. Operation of the WWR-S Reactor in Poland and in the Period 1961-1962; DOKLAD OB ISPOL'ZOVANII I EKSPLUATATSII REAKTORA TIPA VVR-C V POL'SHE ZA 1961-1962 GODA

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

    Aleksandrowicz, J.

    1963-03-01

    The experimental equipment used in the work at the horizontal reactor channels is listed. Diagrams of the utilization of the nominal reactor power and core loading are given, the reactivity fractions of the separate fuel assemblies are detonated, together with the diagram of reactivity versus burnup. Reactor channels and space used for sample irradiation and isotope production are described, and the total number of irradiations is given. Results of the measurements connected with the routine reactor operation are quoted, namely: analysis of water purity in the primary circuit, analysis of the work of the ion exchanger and mechanical filter, andmore » analysis of air activity in the special ventilation system. Data are given concerning radiation protection of personnel, including individual monitoring. Leak testing of the fuel elements is discussed. Damage of the reactor equipment and appearance of alarm signals are described. (auth)« less

  12. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

  13. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

  14. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

  15. Droplet Core Nuclear Rocket (DCNR)

    NASA Technical Reports Server (NTRS)

    Anghaie, Samim

    1991-01-01

    The most basic design feature of the droplet core nuclear reactor is to spray liquid uranium into the core in the form of droplets on the order of five to ten microns in size, to bring the reactor to critical conditions. The liquid uranium fuel ejector is driven by hydrogen, and more hydrogen is injected from the side of the reactor to about one and a half meters from the top. High temperature hydrogen is expanded through a nozzle to produce thrust. The hydrogen pressure in the system can be somewhere between 50 and 500 atmospheres; the higher pressure is more desirable. In the lower core region, hydrogen is tangentially injected to serve two purposes: (1) to provide a swirling flow to protect the wall from impingement of hot uranium droplets: (2) to generate a vortex flow that can be used for fuel separation. The reactor is designed to maximize the energy generation in the upper region of the core. The system can result in and Isp of 2000 per second, and a thrust-to-weight ratio of 1.6 for the shielded reactor. The nuclear engine system can reduce the Mars mission duration to less than 200 days. It can reduce the hydrogen consumption by a factor of 2 to 3, which reduces the hydrogen load by about 130 to 150 metric tons.

  16. 40 CFR 63.1406 - Reactor batch process vent provisions.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

  17. 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...

  18. 40 CFR 63.1406 - Reactor batch process vent provisions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

  19. Layer Protecting the Surface of Zirconium Used in Nuclear Reactors.

    PubMed

    Ashcheulov, Petr; Skoda, Radek; Skarohlíd, Jan; Taylor, Andrew; Fendrych, Frantisek; Kratochvílová, Irena

    2016-01-01

    Zirconium alloys have very useful properties for nuclear facilities applications having low absorption cross-section of thermal electrons, high ductility, hardness and corrosion resistance. However, there is also a significant disadvantage: it reacts with water steam and during this (oxidative) reaction it releases hydrogen gas, which partly diffuses into the alloy forming zirconium hydrides. A new strategy for surface protection of zirconium alloys against undesirable oxidation in nuclear reactors by polycrystalline diamond film has been patented- Czech patent 305059: Layer protecting the surface of zirconium alloys used in nuclear reactors and PCT patent: Layer for protecting surface of zirconium alloys (Patent Number: WO2015039636-A1). The zirconium alloy surface was covered by polycrystalline diamond layer grown in plasma enhanced chemical vapor deposition apparatus with linear antenna delivery system. Substantial progress in the description and understanding of the polycrystalline diamond/ zirconium alloys interface and material properties under standard and nuclear reactors conditions (irradiation, hot steam oxidation experiments and heating-quenching cycles) was made. In addition, process technology for the deposition of protective polycrystalline diamond films onto the surface of zirconium alloys was optimized. Zircaloy2 nuclear fuel pins were covered by 300 nm thick protective polycrystalline diamond layer (PCD) using plasma enhanced chemical vapor deposition apparatus with linear antenna delivery system. The polycrystalline diamond layer protects the zirconium alloy surface against undesirable oxidation and consolidates its chemical stability while preserving its functionality. PCD covered Zircaloy2 and standard Zircaloy2 pins were for 30 min. oxidized in 1100°C hot steam. Under these conditions α phase of zirconium changes to β phase (more opened for oxygen/hydrogen diffusion). PCD anticorrosion protection of Zircaloy nuclear fuel assemblies can significantly prolong lifetime of Zirconium alloy in nuclear reactors even above Zirconium phase transition temperatures. Even after ion beam irradiation (10 dpa, 3 MeV Fe(2+)) the diamond film still shows satisfactory structural integrity with both sp(3) and sp(2) carbon phases. Zircaloy2 under the carbon-based protective layer after hot steam oxidation test differed from the original Zircaloy2 material composition only very slightly, proving that the diamond coating increases the material resistance to high temperature oxidation. Zirconium alloys nuclear fuel pins' surfaces were covered by compact and homogeneous polycrystalline diamond layers consisting of sp(3) and sp(2) carbon phases with a high crystalline diamond content and low roughness. Diamond withstands very high temperatures, has excellent thermal conductivity and low chemical reactivity, it does not degrade over time and (important for the nuclear fuel cladding) being pure carbon, it has perfect neutron cross-section properties. Moreover, polycrystalline diamond layers consisting of crystalline (sp(3)) and amorphous (sp(2)) carbon phases could have suitable thermal expansion. Zirconium alloys coated with polycrystalline diamond film are protected against undesirable changes and processes. Further, the polycrystalline diamond layer prevents the reaction between the alloy surface and water vapor. During such reaction, water molecules dissociate and initiate formation of zirconium dioxide and hydrogen, accompanied by the release of large amount of heat. Thus the protective layer prevents the formation of hydrogen and the release of reaction heat. Few relevant patents to the topic have been reviewed and cited.

  20. Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

    NASA Astrophysics Data System (ADS)

    Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

    2018-02-01

    RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

  1. Game theoretic analysis of physical protection system design

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

    Canion, B.; Schneider, E.; Bickel, E.

    The physical protection system (PPS) of a fictional small modular reactor (SMR) facility have been modeled as a platform for a game theoretic approach to security decision analysis. To demonstrate the game theoretic approach, a rational adversary with complete knowledge of the facility has been modeled attempting a sabotage attack. The adversary adjusts his decisions in response to investments made by the defender to enhance the security measures. This can lead to a conservative physical protection system design. Since defender upgrades were limited by a budget, cost benefit analysis may be conducted upon security upgrades. One approach to cost benefitmore » analysis is the efficient frontier, which depicts the reduction in expected consequence per incremental increase in the security budget.« less

  2. Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

    NASA Astrophysics Data System (ADS)

    Permana, Sidik

    2017-07-01

    Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

  3. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

  4. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

  5. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

  6. Nuclear Technology Series. Course 9: Reactor Auxiliary Systems.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  7. Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

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

    Stroh, K.R.

    1980-01-01

    The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

  8. Oregon State University TRIGA Reactor annual report

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

    Anderson, T.V.; Johnson, A.G.; Bennett, S.L.

    1979-08-31

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included.

  9. 75 FR 9626 - Notice of Issuance of Regulatory Guide

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-03

    ... maintain variables and systems that can affect the fission process, the integrity of the reactor core, the... on some other defined basis. GDC 24 requires that interconnection of the protection and control..., Regulatory Guide Development Branch, Division of Engineering, Office of Nuclear Regulatory Research. [FR Doc...

  10. Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

    NASA Astrophysics Data System (ADS)

    Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

  11. 10 CFR 50.36 - Technical specifications.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

  12. 10 CFR 50.36 - Technical specifications.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

  13. 10 CFR 50.36 - Technical specifications.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

  14. 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...

  15. 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...

  16. 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...

  17. 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...

  18. 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...

  19. Westinghouse Small Modular Reactor balance of plant and supporting systems design

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

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operationmore » of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)« less

  20. Nuclear Technology Series. Course 6: Instrumentation and Control of Reactors and Plant Systems.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  1. Thermal Stratification Analysis for Sodium Fast Reactors

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

    Schneider, James; Anderson, Mark; Baglietto, Emilio

    The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

  2. Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

    NASA Astrophysics Data System (ADS)

    Lee, Chang Jae; Yun, Jae Hee

    2017-06-01

    Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

  3. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

  4. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

  5. Comprehensive work plan for Building 3001 storage canal at the Oak Ridge National Laboratory, Oak Ridge, Tennessee

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

    NONE

    1997-01-01

    This Comprehensive Work Plan describes the method of accomplishment to replace the shielding protection of the water in the canal with a controlled low strength material (CLSM) 4. The canal was used during the operation of the Oak Ridge Graphite Reactor in the 1940s and 1950s to transport spent fuel slugs and irradiated test materials from the reactor, under water to the hot cell in Building 3019 for further processing, packaging, and handling. After the reactor was shut down, the canal was used until 1990 to store some irradiated materials until they could be transferred to a Solid Waste Storagemore » Area. This task has the following objectives and components: (1) minimize potential future risk to human health and the environment; (2) reduce surveillance and maintenance cost of the canal; (3) perform site preparation activities; (4) replace the water in the canal with a solid CLSM; (5) pump the water to the Process Waste Treatment System (PWTS) for further processing at the same rate that the CLSM is pumped under the water; (6) remove the water using a process that will protect the workers and the public in the visitors area from contamination while the CLSM is being pumped underneath the water; (7) painting a protective coating material over the CLSM after the CLSM has cured.« less

  6. 77 FR 68161 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-15

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on December 4, 2012, Room T-2B3, 11545...

  7. 77 FR 31044 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-24

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on June 5, 2012, Room T-2B1, 11545 Rockville...

  8. 77 FR 56240 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-12

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on September 18, 2012, Room T-2B3, 11545...

  9. 76 FR 44964 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-27

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on August 17, 2011, Room T-2B3, 11545...

  10. 78 FR 70597 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-26

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on December 3, 2013, Room T-2B1, 11545...

  11. 78 FR 17944 - Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-25

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on April 9, 2013, Room T-2B3, 11545 Rockville...

  12. 77 FR 38099 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-26

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on July 10, 2012, Room T-2B1, 11545 Rockville...

  13. 78 FR 66967 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommitte on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-07

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommitte on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on November 19, 2013, Room T-2B1, 11545...

  14. 76 FR 55717 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-08

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on September 23, 2011, Room T-2B3, 11545...

  15. 75 FR 27840 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-18

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on May 18, 2010, Room T-2B1, 11545 Rockville Pike, Rockville...

  16. 75 FR 27841 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-18

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on May 18, 2010, Room T-2B1, 11545 Rockville Pike, Rockville...

  17. 76 FR 27101 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-10

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on May 25, 2011, Room T-2B3, 11545 Rockville...

  18. 76 FR 34779 - Advisory Committee On Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-14

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on June 23, 2011, Room T-2B1, 11545 Rockville...

  19. 75 FR 58447 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-24

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on October 22, 2010, Room T-2B1, 11545 Rockville Pike, Rockville...

  20. 75 FR 16874 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-04-02

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on April 21, 2010, Room T2-B3, at 11545...

  1. 76 FR 55716 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of The ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-08

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of The ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on September 22, 2011, Room T-2B1, 11545...

  2. 75 FR 82093 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-29

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on January 11, 2011, Room T-2B3, 11545...

  3. 75 FR 4881 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-01-29

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on February 17, 2010, Room T2-B1, 11545...

  4. 75 FR 82092 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-29

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on January 12, 2011, Room T-2B3, 11545...

  5. 78 FR 79020 - Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-27

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on January 16, 2014, Room T-2B1, 11545...

  6. 76 FR 61119 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-03

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on October 4, 2011, Room T-2B1, 11545...

  7. Magnet design with 100-kA HTS STARS conductors for the helical fusion reactor

    NASA Astrophysics Data System (ADS)

    Yanagi, N.; Terazaki, Y.; Ito, S.; Tamura, H.; Hamaguchi, S.; Mito, T.; Hashizume, H.; Sagara, A.

    2016-12-01

    The high-temperature superconducting (HTS) option is employed for the conceptual design of the LHD-type helical fusion reactor FFHR-d1. The 100-kA-class STARS (Stacked Tapes Assembled in Rigid Structure) conductor is used for the magnet system including the continuously wound helical coils. Protection of the magnet system in case of a quench is a crucial issue and the hot-spot temperature during an emergency discharge is estimated based on the zero-dimensional and one-dimensional analyses. The number of division of the coil winding package is examined to limit the voltage generation. For cooling the HTS magnet, helium gas flow is considered and its feasibility is examined by simple analysis as a first step.

  8. Hydrogen-Enhanced Lunar Oxygen Extraction and Storage Using Only Solar Power

    NASA Technical Reports Server (NTRS)

    Burton, rodney; King, Darren

    2013-01-01

    The innovation consists of a thermodynamic system for extracting in situ oxygen vapor from lunar regolith using a solar photovoltaic power source in a reactor, a method for thermally insulating the reactor, a method for protecting the reactor internal components from oxidation by the extracted oxygen, a method for removing unwanted chemical species produced in the reactor from the oxygen vapor, a method for passively storing the oxygen, and a method for releasing high-purity oxygen from storage for lunar use. Lunar oxygen exists in various types of minerals, mostly silicates. The energy required to extract the oxygen from the minerals is 30 to 60 MJ/kg O. Using simple heating, the extraction rate depends on temperature. The minimum temperature is approximately 2,500 K, which is at the upper end of available oven temperatures. The oxygen is released from storage in a purified state, as needed, especially if for human consumption. This method extracts oxygen from regolith by treating the problem as a closed batch cycle system. The innovation works equally well in Earth or Lunar gravity fields, at low partial pressure of oxygen, and makes use of in situ regolith for system insulation. The innovation extracts oxygen from lunar regolith using a method similar to vacuum pyrolysis, but with hydrogen cover gas added stoichiometrically to react with the oxygen as it is produced by radiatively heating regolith to 2,500 K. The hydrogen flows over and through the heating element (HE), protecting it from released oxygen. The H2 O2 heat of reaction is regeneratively recovered to assist the heating process. Lunar regolith is loaded into a large-diameter, low-height pancake reactor powered by photovoltaic cells. The reactor lid contains a 2,500 K HE that radiates downward onto the regolith to heat it and extract oxygen, and is shielded above by a multi-layer tungsten radiation shield. Hydrogen cover gas percolates through the perforated tungsten shielding and HE, preventing oxidation of the shielding and HE, and reacting with the oxygen to form water vapor. The water vapor is filtered through solid regolith to remove unwanted extraction byproducts, and then condensed to a liquid state and stored at 300 to 325 K. Conversion to usable oxygen is achieved by pumping liquid water into a high-pressure electrolyzer, storing the gaseous oxygen at high pressure for use, and diverting the hydrogen back to the reactor or to storage. The results from this design effort show that this oxygen-generating concept can be developed in an efficient system with low specific mass. Advantages include use of regolith as an oxygen source, filter, and thermal insulator. The system can be tested in Earth gravity and can be expected to operate similarly in lunar gravity. The system is scalable, either by increasing the power level and output of a standard module, or by employing multiple modules.

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

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as thesemore » installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.« less

  10. Furfural production from biomass pretreatment hydrolysate using vapor-releasing reactor system.

    PubMed

    Liu, Lu; Chang, Hou-Min; Jameel, Hasan; Park, Sunkyu

    2018-03-01

    Biomass hydrolysate from autohydrolysis pretreatment was used for furfural production considering it is in rich of xylose, xylo-oligomers, and other decomposition products from hemicellulose structure. By using the vapor-releasing reactor system, furfural was protected from degradation by separating it from the reaction media. The maximum furfural yield of 73% was achieved at 200 °C for biomass hydrolysate without the use of the catalyst. This is because the presence of organic acids such as acetic acid in hydrolysate functioned as a catalyst. According to the results in this study, biomass hydrolysate with a vapor-releasing system proves to be efficient for furfural production. The biorefinery process which allows the separation of xylose-rich autohydrolysate from other parts from biomass feedstock also improves the overall application of the biomass. Copyright © 2018 Elsevier Ltd. All rights reserved.

  11. Safety survey report EBR-II safety survey, ANL-west health protection, industrial safety and fire protection survey

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

    Dunbar, K.A.

    1972-01-10

    A safety survey covering the disciplines of Reactor Safety, Nuclear Criticality Safety, Health Protection and Industrial Safety and Fire Protection was conducted at the ANL-West EBR-II FEF Complex during the period January 10-18, 1972. In addition, the entire ANL-West site was surveyed for Health Protection and Industrial Safety and Fire Protection. The survey was conducted by members of the AEC Chicago Operations Office, a member of RDT-HQ and a member of the RDT-ID site office. Eighteen recommendations resulted from the survey, eleven in the area of Industrial Safety and Fire Protection, five in the area of Reactor Safety and twomore » in the area of Nuclear Criticality Safety.« less

  12. 76 FR 36160 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-21

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Radiation Protection and Nuclear Materials Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on June 20, 2011, Room T-2B1, 11545 Rockville Pike...

  13. 75 FR 38564 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-02

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations and Fire Protection The ACRS Subcommittee on Plant Operations and Fire Protection will hold a meeting on July 29, 2010, at the U.S. NRC Region IV, Texas Health Resources Tower, 612...

  14. 10 CFR 140.13 - Amount of financial protection required of certain holders of construction permits and combined...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ..., possession and storage only of special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of either an operating license under 10 CFR part 50 or... NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS...

  15. 10 CFR 140.13 - Amount of financial protection required of certain holders of construction permits and combined...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ..., possession and storage only of special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of either an operating license under 10 CFR part 50 or... NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS...

  16. 10 CFR 140.13 - Amount of financial protection required of certain holders of construction permits and combined...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ..., possession and storage only of special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of either an operating license under 10 CFR part 50 or... NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS...

  17. 10 CFR 140.13 - Amount of financial protection required of certain holders of construction permits and combined...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ..., possession and storage only of special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of either an operating license under 10 CFR part 50 or... NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS...

  18. 10 CFR 140.13 - Amount of financial protection required of certain holders of construction permits and combined...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ..., possession and storage only of special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of either an operating license under 10 CFR part 50 or... NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS...

  19. LMFBR system-wide transient analysis: the state of the art and US validation needs

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

    Khatib-Rahbar, M.; Guppy, J.G.; Cerbone, R.J.

    1982-01-01

    This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants.

  20. SETS. Set Equation Transformation System

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

    Worrell, R.B.

    1992-01-13

    SETS is used for symbolic manipulation of Boolean equations, particularly the reduction of equations by the application of Boolean identities. It is a flexible and efficient tool for performing probabilistic risk analysis (PRA), vital area analysis, and common cause analysis. The equation manipulation capabilities of SETS can also be used to analyze noncoherent fault trees and determine prime implicants of Boolean functions, to verify circuit design implementation, to determine minimum cost fire protection requirements for nuclear reactor plants, to obtain solutions to combinatorial optimization problems with Boolean constraints, and to determine the susceptibility of a facility to unauthorized access throughmore » nullification of sensors in its protection system.« less

  1. Evaluation of the Destruction of the Harmful Cyanobacteria, Microcystis aeruginosa, with a Cavitation and Superoxide Generating Water Treatment Reactor.

    PubMed

    Medina, Victor F; Griggs, Chris S; Thomas, Catherine

    2016-06-01

    Cyanobacterial/Harmful Algal Blooms are a major issue for lakes and reservoirs throughout the U.S.A. An effective destructive technology could be useful to protect sensitive areas, such as areas near water intakes. The study presented in this article explored the use of a reactor called the KRIA Water Treatment System. The reactor focuses on the injection of superoxide (O2 (-)), which is generated electrochemically from the atmosphere, into the water body. In addition, the injection process generates a significant amount of cavitation. The treatment process was tested in 190-L reactors spiked with water from cyanobacterial contaminated lakes. The treatment was very effective at destroying the predominant species of cyanobacteria, Microcystis aeruginosa, organic matter, and decreasing chlorophyll concentration. Microcystin toxin concentrations were also reduced. Data suggest that cavitation alone was an effective treatment, but the addition of superoxide improved performance, particularly regarding removal of cyanobacteria and reduction of microcystin concentration.

  2. Enhancing the performance of sequencing batch reactors by adding crushed date seeds to remove high concentrations of 2,4-dinitrophenol.

    PubMed

    Al-Mutairi, Nayef Z

    2011-11-01

    Wastewater treatment systems using simultaneous adsorption and biodegradation processes have been successful in treating toxic pollutants present in industrial wastewater. The goal of this investigation was to assess the effectiveness of date seeds in reducing the toxic effects of 2,4-dinitrophenol (DNP) on activated sludge microorganisms. Two identical sequencing batch reactors (SBRs) (4-L glass vessel), each with a 3.5-L working volume, were used. The initial DNP concentrations in the reactor were 50, 75, 100, 250, and 500 mg/L. The reactor amended with date seeds was capable of degrading DNP at significantly greater rates (11 +/- 2.5 mg/L x h) than the control SBR (4 +/- 1.2 mg/L x h) at a 95% confidence level. Date seeds can be added to the mixed liquor of activated sludge treatment plants to remove high concentrations of DNP from wastewater, to protect the treatment plant against toxic components in the influent and enhance the settling characteristics of the mixed liquor.

  3. A probabilistic safety analysis of incidents in nuclear research reactors.

    PubMed

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  4. NUCLEAR NEW BUILD-INTEGRATING CULTURAL DIFFERENCES IN RADIATION PROTECTION.

    PubMed

    Haemmerli, Valentin; Bryant, Peter A; Cole, Peter

    2017-04-01

    Across the world, we are seeing a resurgence in Nuclear New Build. In the UK alone, plans are under way for the construction of 10 new reactors, using 4 different reactor designs all of which are to be provided by foreign vendors, and operated by 3 newly formed licensees within the UK. As these new licensees embark on the task of establishing themselves and progressing the design and build of these reactors, there are challenges faced in integrating the Radiation Protection Requirements and Culture from the various Foreign Investors and Vendors into the UK 'Context'. The following paper identifies the origin of the Radiation Protection Requirements within the UK and foreign investor/vendor countries, in an attempt to integrate them into the UK licensing and approval process. Thus, allowing due credit to be taken for the regulatory regime of the foreign countries where these reactors originate. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  5. COATED CARBON ELEMENT FOR USE IN NUCLEAR REACTORS AND THE PROCESS OF MAKING THE ELEMENT

    DOEpatents

    Pyle, R.J.; Allen, G.L.

    1963-01-15

    S>This patent relates to a carbide-nitride-carbide coating for carbon bodies that are to be subjected to a high temperature nuclear reactor atmosphere, and a method of applying the same. This coating is a highly efficient diffusion barrier and protects the C body from corrosion and erosion by the reactor atmosphere. Preferably, the innermost coating is Zr carbide, the middle coatlng is Zr nitride, and the outermost coating is a mixture of Zr and Nb carbide. The nitride coating acts as a diffusion barrier, while the innermost carbide bonds the nitride to the C body and prevents deleterious reaction between the nitride and C body. The outermost carbide coating protects the nitride coating from the reactor atmosphere. (AEC)

  6. Evaluation of an Innovative Approach to Validation of ...

    EPA Pesticide Factsheets

    UV disinfection is an effective process for inactivating many microbial pathogens found in source waters with the potential as stand-alone treatment or in combination with other disinfectants. For surface and groundwater sourced drinking water applications, the U.S. Environmental Protection Agency (USEPA) provided guidance on the validation of UV reactors nearly a decade ago. The focus of the guidance was primarily for inactivation of Cryptosporidium and Giardia. Over the last ten years many lessons have been learned, validation practices have been modified, new science issues discovered, and changes in operation & monitoring of UV systems need to be addressed. Also, there remains no standard approach for validating UV reactors to meet a 4-log (99.99%) inactivation of viruses. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater & surface water pathogen inactivation including viruses for low-pressure and medium-pressure UV systems. A particular challenge for medium-pressure UV is the monitoring of low-wavelength germicidal contributions for appropriate crediting of disinfection under varying reactor conditions of quartz sleeve fouling, lamp aging, and changes in UV absorbance of the water over time. In the current effort, bench and full-scale studies are being conducted on a low pressure (LP) UV reactor and a medium pressure (MP) UV re

  7. Generating Breathable Air Through Dissociation of N2O

    NASA Technical Reports Server (NTRS)

    Zubrin, Robert; Frankie, Brian

    2006-01-01

    A nitrous oxide-based oxygen-supply system (NOBOSS) is an apparatus in which a breathable mixture comprising 2/3 volume parts of N2 and 1/3 volume part of O2 is generated through dissociation of N2O. The NOBOSS concept can be adapted to a variety of applications in which there are requirements for relatively compact, lightweight systems to supply breathable air. These could include air-supply systems for firefighters, divers, astronauts, and workers who must be protected against biological and chemical hazards. A NOBOSS stands in contrast to compressed-gas and cryogenic air-supply systems. Compressed-gas systems necessarily include massive tanks that can hold only relatively small amounts of gases. Alternatively, gases can be stored compactly in greater quantities and at low pressures when they are liquefied, but then cryogenic equipment is needed to maintain them in liquid form. Overcoming the disadvantages of both compressed-gas and cryogenic systems, the NOBOSS exploits the fact that N2O can be stored in liquid form at room temperature and moderate pressure. The mass of N2O that can be stored in a tank of a given mass is about 20 times the mass of compressed air that can be stored in a tank of equal mass. In a NOBOSS, N2O is exothermically dissociated to N2 and O2 in a main catalytic reactor. In order to ensure the dissociation of N2O to the maximum possible extent, the temperature of the reactor must be kept above 400 C. At the same time, to minimize concentrations of nitrogen oxides (which are toxic), it is necessary to keep the reactor temperature at or below 540 C. To keep the temperature within the required range throughout the reactor and, in particular, to prevent the formation of hot spots that would be generated by local concentrations of the exothermic dissociation reaction, the N2O is introduced into the reactor through an injector tube that features carefully spaced holes to distribute the input flow of N2O widely throughout the reactor. A NOBOSS includes one or more "destroyer" subsystems for removing any nitrogen oxides that remain downstream of the main N2O-dissociation reactor. A destroyer includes a carbon bed in series with a catalytic reactor, and is in thermal contact with the main N2O-dissociation reactor. The gas mixture that leaves the main reactor first goes through a carbon bed, which adsorbs all of the trace NO and most of the trace NO2. The gas mixture then goes through the destroyer catalytic reactor, wherein most or all of the remaining NO2 is dissociated. A NOBOSS can be designed to regulate its reactor temperature across a range of flow rates. One such system includes three destroyer loops; these loops act, in combination with a heat sink, to remove heat from the main N2O-dissociation reactor. In this system, the N2O and product gases play an additional role as coolants; thus, as needed, the coolant flow increases in proportion to the rate of generation of heat, helping to keep the main-reactor temperature below 540 C.

  8. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

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

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

  9. Wastewater infrastructure for small cities in an urbanizing world: integrating protection of human health and the environment with resource recovery and food security.

    PubMed

    Verbyla, Matthew E; Oakley, Stewart M; Mihelcic, James R

    2013-04-16

    The majority of population growth in developing countries will occur in small cities closely linked to agricultural zones, with poor access to water and sanitation. Wastewater management priorities in these regions will be different from those in larger cities and developed countries. Two wastewater treatment systems in Bolivia, one with an upflow anaerobic sludge blanket (UASB) reactor and polishing ponds, the other with three stabilization ponds, are assessed to determine their resource recovery potential. The UASB reactor produces biogas with 500-650 MJ per day. In six months, both systems discharge wastewater with the same mass of nutrients as fertilizers used to produce crops containing 10-75 days' worth of the recommended food energy intake for each person using the system. Both systems also discharge detectable levels of helminth eggs, Giardia cysts, and Cryptosporidium oocysts, but the UASB reactor system discharges higher concentrations, implying limited reuse potential. From a regional management standpoint, small cities should not expend resources to treat wastewater to levels suitable for discharge into surface waters. Rather, they should focus on removing pathogens to reclaim water and nutrients. Biogas recovery may be a priority that should be subservient to water and nutrient recovery in these settings.

  10. SPERTI. Detail view of Reactor Pit Building (PER605) and Instrument ...

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

    SPERT-I. Detail view of Reactor Pit Building (PER-605) and Instrument Cell (PER-606). Earth shielding covers side of Cell Building next to reactor. Instrumentation required protection from radiation emitted during reactor operation. Photographer: R.G. Larsen. Date: May 20, 1955. INEEL negative no. 55-1290 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  11. Architecture for nuclear energy in the 21st century

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

    Arthu, E.D.; Cunningham, P.T.; Wagner, R.L. Jr.

    1999-02-21

    Global and regional scenarios for future energy demand have been assessed from the perspectives of nuclear materials management. From these the authors propose creation of a nuclear fuel cycle architecture which maximizes inherent protection of plutonium and other nuclear materials. The concept also provides technical and institutional flexibility for transition into other fuel cycle systems, particularly those involving breeder reactors. The system, its implementation timeline, and overall impact are described in the paper.

  12. Anammox enrichment from reject water on blank biofilm carriers and carriers containing nitrifying biomass: operation of two moving bed biofilm reactors (MBBR).

    PubMed

    Zekker, Ivar; Rikmann, Ergo; Tenno, Toomas; Lemmiksoo, Vallo; Menert, Anne; Loorits, Liis; Vabamäe, Priit; Tomingas, Martin; Tenno, Taavo

    2012-07-01

    The anammox bacteria were enriched from reject water of anaerobic digestion of municipal wastewater sludge onto moving bed biofilm reactor (MBBR) system carriers-the ones initially containing no biomass (MBBR1) as well as the ones containing nitrifying biomass (MBBR2). Duration of start-up periods of the both reactors was similar (about 100 days), but stable total nitrogen (TN) removal efficiency occurred earlier in the system containing nitrifying biomass. Anammox TN removal efficiency of 70% was achieved by 180 days in both 20 l volume reactors at moderate temperature of 26.0°C. During the steady state phase of operation of MBBRs the average TN removal efficiencies and maximum TN removal rates in MBBR1 were 80% (1,000 g-N/m(3)/day, achieved by 308 days) and in MBBR2 85% (1,100 g-N/m(3)/day, achieved by 266 days). In both reactors mixed bacterial cultures were detected. Uncultured Planctomycetales bacterium clone P4, Candidatus Nitrospira defluvii and uncultured Nitrospira sp. clone 53 were identified by PCR-DGGE from the system initially containing blank biofilm carriers as well as from the nitrifying biofilm system; from the latter in addition to these also uncultured ammonium oxidizing bacterium clone W1 and Nitrospira sp. clone S1-62 were detected. FISH analysis revealed that anammox microorganisms were located in clusters in the biofilm. Using previously grown nitrifying biofilm matrix for anammox enrichment has some benefits over starting up the process from zero, such as less time for enrichment and protection against severe inhibitions in case of high substrate loading rates.

  13. Interior of the Plum Brook Reactor Facility

    NASA Image and Video Library

    1961-02-21

    A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

  14. Optimizing the switching time for 400 kV SF6 circuit breakers

    NASA Astrophysics Data System (ADS)

    Ciulica, D.

    2018-01-01

    This paper presents real-time voltage and current analysis for optimizing the wave switching point of the circuit breaker SF6. Circuit Breaker plays an important role in power systems. It provides protection for equipment in embedded stations in transport networks. SF6 Circuit Breaker is very important equipment in Power Systems, which is used for up to 400 kV due to its excellent performance. The controlled switching is used to eliminate transient modes and electrodynamic and dielectric charges in the network at manual switching of capacitor, shunt reactors and power transformers. These effects reduce the reliability and lifetime of the equipment installed on the network, or may lead to erroneous protection.

  15. 10 CFR 73.72 - Requirement for advance notice of shipment of formula quantities of strategic special nuclear...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ..., or irradiated reactor fuel. 73.72 Section 73.72 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... strategic significance, or irradiated reactor fuel. (a) A licensee, other than one specified in paragraph (b... strategic significance, or irradiated reactor fuel required to be protected in accordance with § 73.37...

  16. 10 CFR 73.72 - Requirement for advance notice of shipment of formula quantities of strategic special nuclear...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ..., or irradiated reactor fuel. 73.72 Section 73.72 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED... strategic significance, or irradiated reactor fuel. (a) A licensee, other than one specified in paragraph (b... strategic significance, or irradiated reactor fuel required to be protected in accordance with § 73.37...

  17. Protection of tokamak plasma facing components by a capillary porous system with lithium

    NASA Astrophysics Data System (ADS)

    Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

    2015-08-01

    Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

  18. The Euratom Seventh Framework Programme FP7 (2007-2011)

    NASA Astrophysics Data System (ADS)

    Garbil, R.

    2010-10-01

    The objective of the Seventh Euratom Framework Program in the area of nuclear fission and radiation protection is to establish a sound scientific and technical basis to accelerate practical developments of nuclear energy related to resource efficiency, enhancing safety performance, cost-effectiveness and safer management of long-lived radioactive waste. Key cross-cutting topics such as the nuclear fuel cycle, actinide chemistry, risk analysis, safety assessment, even societal and governance issues are linked to the individual technical areas. Research need to explore new scientific and techno- logical opportunities and to respond in a flexible way to new policy needs that arise. The following activities are to be pursued. (a) Management of radioactive waste, research on partitioning and transmutation and/or other concepts aimed at reducing the amount and/or hazard of the waste for disposal; (b) Reactor systems research to underpin the con- tinued safe operation of all relevant types of existing reactor systems (including fuel cycle facilities), life-time extension, development of new advanced safety assessment methodologies and waste-management aspects of future reactor systems; (c) Radiation protection research in particular on the risks from low doses on medical uses and on the management of accidents; (d) Infrastructures and support given to the availability of, and cooperation between, research infrastructures necessary to maintain high standards of technical achievement, innovation and safety in the European nuclear sector and Research Area. (e) Human resources, mobility and training support to be provided for the retention and further development of scientific competence, human capacity through joint training activities in order to guarantee the availability of suitably qualified researchers, engineers and employees in the nuclear sector over the longer term.

  19. 10 CFR 73.4 - Communications.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...: Document Control Desk, Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors...://www.nrc.gov/site-help/e-submittals.html; by e-mail to [email protected]; or by writing the Office...

  20. Review of Rover fuel element protective coating development at Los Alamos

    NASA Technical Reports Server (NTRS)

    Wallace, Terry C.

    1991-01-01

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

  1. Experience with chemical system decontamination by the CORD process and electrochemical decontamination of pipe ends

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

    Wille, H.; Bertholdt, H.O.; Operschall, H.

    Efforts to reduce occupational radiation exposure during inspection and repair work in nuclear power plants turns steadily increasing attention to the decontamination of systems and components. Due to the advanced age of nuclear power plants resulting in increasing dose rates, the decontamination of components, or rather of complete systems, or loops to protect operating and inspection personnel becomes demanding. Besides, decontaminating complete primary loops is in many cases less difficult than cleaning large components. Based on experience gained in nuclear power plants, an outline of two different decontamination methods performed recently are given. For the decontamination of complete systems ormore » loops, Kraftwerk Union AG has developed CORD, a low-concentration process. For the decontamination performance of a subsystem, such as the steam generator (SG) channel heads of a pressurized water reactor or the recirculation loops of a boiling water reactor the automated mobile decontamination appliance is used. The electrochemical decontamination process is primarily applicable for the treatment of specially limited surface areas.« less

  2. TECHNICAL AND MEDICAL ASPECTS OF NUCLAR-POWERED AIRPLANES: THE ROLE OF NUCLEAR POWER IN THE ATMOSPHERE AND IN SPACE (in Swedish)

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

    von Ublisch, H.

    1961-01-01

    An evaluation is given of the merits of potential nuclear propulsion systems for aircraft and rockets as well as of small nuclear power plants for auxiliary use in space exploration. The protection of personnel and passengers against gamma and high-velocity radiation is discussed, and the shielding properties of various materials are analysed. Mobile shielding would be required for airplanes even when landed and the reactor is shut down. No significant pollutton of the atmosphere is expected from leaking reactors, but accidents constitute a real danger. The prospects of realizing the ion motor and the photon motor are speculated upon. (auth)

  3. The investigation of a tuberculosis outbreak in the closed environment of a U.S. Navy ship, 1987.

    PubMed

    DiStasio, A J; Trump, D H

    1990-08-01

    A sailor on a U.S. Navy ship had smear-positive, cavitary, pulmonary tuberculosis. Contact investigation of the entire ship's crew found 216 new reactors to tuberculin skin test (24.5%) among 881 previously tuberculin-negative sailors. The risk for new infection was highest among sailors in the patient's department (relative risk, 4.4; 95% confidence interval 3.7, 5.3); 95% (15/16) of sailors in his division were new reactors. While crewmembers in all departments were at risk for a new tuberculosis infection, working and berthing in compartments that were distant from those of the index case were protective. The ship's closed ventilation system contributed to the outbreak.

  4. Protective interior wall and attach8ing means for a fusion reactor vacuum vessel

    DOEpatents

    Phelps, Richard D.; Upham, Gerald A.; Anderson, Paul M.

    1988-01-01

    An array of connected plates mounted on the inside wall of the vacuum vessel of a magnetic confinement reactor in order to provide a protective surface for energy deposition inside the vessel. All fasteners are concealed and protected beneath the plates, while the plates themselves share common mounting points. The entire array is installed with torqued nuts on threaded studs; provision also exists for thermal expansion by mounting each plate with two of its four mounts captured in an oversize grooved spool. A spool-washer mounting hardware allows one edge of a protective plate to be torqued while the other side remains loose, by simply inverting the spool-washer hardware.

  5. Impact of Active Control on Passive Safety Response Characteristics of Sodium-Cooled Fast Reactors: II-Model Implementation and Simulations

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

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    Advanced reactors are often claimed to be passively safe against unprotected upset events. In common practice, these events are not considered in the context of the plant control system, i.e., the reactor is subjected to classes of unprotected upset events while the normally programmed response of the control system is assumed not to be present. However, this approach constitutes an oversimplification since, depending on the upset involving the control system, an actuator does not necessarily go in the same direction as needed for safety. In this work, dynamic simulations are performed to assess the degree to which the inherent self-regulatingmore » plant response is safe from active control system override. The simulations are meant to characterize the resilience of the plant to unprotected initiators. The initiators were represented and modeled as an actuator going to a hard limit. Consideration of failure is further limited to individual controllers as there is no cross-connect of signals between these controllers. The potential for passive safety override by the control system is then relegated to the single-input single-output controllers. Here, the results show that when the plant control system is designed by taking into account and quantifying the impact of the plant control system on accidental scenarios there is very limited opportunity for the preprogrammed response of the control system to override passive safety protection in the event of an unprotected initiator.« less

  6. Impact of Active Control on Passive Safety Response Characteristics of Sodium-Cooled Fast Reactors: II-Model Implementation and Simulations

    DOE PAGES

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    2017-06-21

    Advanced reactors are often claimed to be passively safe against unprotected upset events. In common practice, these events are not considered in the context of the plant control system, i.e., the reactor is subjected to classes of unprotected upset events while the normally programmed response of the control system is assumed not to be present. However, this approach constitutes an oversimplification since, depending on the upset involving the control system, an actuator does not necessarily go in the same direction as needed for safety. In this work, dynamic simulations are performed to assess the degree to which the inherent self-regulatingmore » plant response is safe from active control system override. The simulations are meant to characterize the resilience of the plant to unprotected initiators. The initiators were represented and modeled as an actuator going to a hard limit. Consideration of failure is further limited to individual controllers as there is no cross-connect of signals between these controllers. The potential for passive safety override by the control system is then relegated to the single-input single-output controllers. Here, the results show that when the plant control system is designed by taking into account and quantifying the impact of the plant control system on accidental scenarios there is very limited opportunity for the preprogrammed response of the control system to override passive safety protection in the event of an unprotected initiator.« less

  7. 75 FR 72653 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-26

    ..., Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors... Requirements for Protection Against Pressurized Thermal Shock Events; Correction AGENCY: Nuclear Regulatory... fracture toughness requirements for protection against pressurized thermal shock (PTS) events for...

  8. Characteristics and Dose Levels for Spent Reactor Fuels

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

    Coates, Cameron W

    2007-01-01

    Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

  9. Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

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

    Wichner, R.P.; Hodge, S.A.; Weber, C.F.

    1984-08-01

    This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

  10. Hanford Laboratories monthly activities report, February 1964

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

    Not Available

    1964-03-16

    This is the monthly report for the Hanford Laboratories Operation, February, 1964. Reactor fuels, chemistry, dosimetry, separation process, reactor technology financial activities, biology operation, physics and instrumentation research, employee relations, applied mathematics, programming, and radiation protection are discussed.

  11. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  12. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  13. Advanced refractory metals and composites for extraterrestrial power systems

    NASA Technical Reports Server (NTRS)

    Titran, R. H.; Grobstein, Toni L.

    1990-01-01

    Concepts for future space power systems include nuclear and focused solar heat sources coupled to static and dynamic power-conversion devices; such systems must be designed for service lives as long as 30 years, despite service temperatures of the order of 1600 K. Materials are a critical technology-development factor in such aspects of these systems as reactor fuel containment, environmental protection, power management, and thermal management. Attention is given to the prospective performance of such refractory metals as Nb, W, and Mo alloys, W fiber-reinforced Nb-matrix composites, and HfC precipitate-strengthened W-Re alloys.

  14. Transuranic inventory reduction in repository by partitioning and transmutation

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

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

    1992-01-01

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

  15. Influence of oxide microstructure on corrosion behavior of zirconium-based model alloys

    NASA Astrophysics Data System (ADS)

    Silva, Marcelo Jose Gomes Da

    The extensive utilization of zirconium-based alloys in fuel cladding and other reactor internal components in the nuclear power industry has led to the continuous improvement of these alloys. At the present moment, demands for better performing nuclear fuel cladding materials are increasing. Also, new reactor designs have been proposed that would require the materials to withstand even more rigorous conditions. One of the factors that limit s fuel cladding utilization in nuclear reactors is uniform corrosion and the consequent hydriding of the fuel. In an attempt to develop mechanistic understanding of the role of alloying elements in the growth of a stable protective oxide, a series of model zirconium-based alloys was prepared (Zr-xFe-yCr, Zr-xCu-yMo, Zr-xNb-ySn, for various x and y, pure Zr and Zircaloy-4) and examined with advanced characterization techniques. The alloys were corrosion tested in autoclaves under three different conditions: 360°C water, 500°C steam and 500°C supercritical water in excess of 400 days. These autoclave testing conditions simulate nuclear reactor environment for both current designs (360°C water) and the new supercritical water reactor (500°C steam and 500°C supercritical water) proposed by the generation-IV initiative. The oxide films formed were systematically examined at the Advanced Photon Source using microbeam synchrotron radiation diffraction and fluorescence of cross-sectional samples to determine the oxide phases present and their crystallographic texture as a function of distance from the metal/oxide interface. Also, the overall texture of the oxide layers was investigated using synchrotron radiation diffraction in frontal geometry. The corrosion kinetics is a function of the alloy system and showed a wide range of behaviors, from immediately unstable oxide growth to stable behavior. The corrosion weight gains from testing at high temperature are a factor of five higher than those measured at 360°C but the protectiveness ranking of the alloys is similar. Measured pole figures from different oxides in different corrosion regimes showed that monoclinic oxides grow in slightly distinct directions: protective oxides grow along the (-904)m pole, whether non-protective oxides grow along or close to the (-302)m pole. The angle in between these two directions ((-904)m and (-302)m) is about 6°. Microbeam synchrotron radiation diffraction and fluorescence was performed in the oxide layers and systematic differences are observed in protective and non-protective oxides, both near the oxide/metal interface and in the bulk of the oxide layers. The non-protective oxide interfaces show a smooth transition from metal to oxide with metal diffraction peaks disappearing as the monoclinic oxide peaks appear. In contrast, in a protective oxide, a complex structure near the oxide/metal interface was seen, showing peaks from Zr 3O suboxide and a highly oriented tetragonal oxide phase with specific orientation relationships with the monoclinic oxide and the base metal. The highly oriented tetragonal phase, only present in protective oxides, is believed to be a precursor to the formation of monoclinic oxide found in the bulk of the oxide layer. This plane may promote stable growth by causing the oxide to form in a manner that maximizes occupation of the substrate surface and minimizes stress accumulation, leading to more stable oxide growth. The association seen in this work of the precursor oxide phase with protective oxides and its orientation relationship with the monoclinic oxide, combined with the difference in oxide growth direction seen between protective and non-protective oxides, is interpreted as evidence that this phase allows a more properly oriented oxide to grow, in a way that minimizes stress accumulation and therefore delays the oxide transition to larger oxide thicknesses.

  16. Summary of the Forty-Fifth NCRP annual meeting on "the future of nuclear power worldwide: safety, health and the environment".

    PubMed

    Corradini, Michael

    2011-01-01

    The role of nuclear power as a major resource in meeting the projected growth of electric power requirements in the United States and worldwide during the 21st century is a subject of great contemporary interest. The goal of the 2009 NCRP Annual Meeting was to provide a forum for an in-depth discussion of issues related to the safety, health and environmental protection aspects of new nuclear power reactor systems and related fuel-cycle facilities such as fuel production and reprocessing strategies. The meeting was an international conference with participation of almost 400 representatives from many nations, scientific organizations, nuclear industries, and governmental agencies engaged in the development and regulatory control of advanced nuclear reactor systems and fuel-cycle operations. Highlights of the meeting are summarized in this report. Copyright © 2010 Health Physics Society

  17. The Potential of the LFR and the ELSY Project

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

    Cinotti, L; Smith, C F; Sienicki, J J

    2007-03-12

    This paper presents the current status of the development of the Lead-cooled Fast Reactor (LFR) in support of Generation IV (GEN IV) Nuclear Energy Systems. The approach being taken by the GIF plan is to address the research priorities of each member state in developing an integrated and coordinated research program to achieve common objectives, while avoiding duplication of effort. The integrated plan being prepared by the LFR Provisional System Steering Committee of the GIF, known as the LFR System research Plan (SRP) recognizes two principal technology tracks for pursuit of LFR technology: (1) a small, transportable system of 10-100more » MWe size that features a very long refueling interval, (2) a larger-sized system rated at about 600 MWe, intended for central station power generation and waste transmutation. This paper, in particular, describes the ongoing activities to develop the Small Secure Transportable Autonomous Reactor (SSTAR) and the European Lead-cooled SYstem (ELSY), the two research initiatives closely aligned with the overall tracks of the SRP and outlines the Proliferation-resistant Environment-friendly Accident-tolerant Continual & Economical Reactors (PEACER) conceived with particular focus on burning/transmuting of long-living TRU waste and fission fragments of concern, such as Tc and I. The current reference design for the SSTAR is a 20 MWe natural circulation pool-type reactor concept with a small shippable reactor vessel. Specific features of the lead coolant, the nitride fuel containing transuranics, the fast spectrum core, and the small size combine to promote a unique approach to achieve proliferation resistance, while also enabling fissile self-sufficiency, autonomous load following, simplicity of operation, reliability, transportability, as well as a high degree of passive safety. Conversion of the core thermal power into electricity at a high plant efficiency of 44% is accomplished utilizing a supercritical carbon dioxide Brayton cycle power converter. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. This concept has been under development since September 2006, and is sponsored by the Sixth Framework Programme of EURATOM. The ELSY project is being performed by a consortium consisting of twenty organizations including seventeen from Europe, two from Korea and one from the USA. ELSY aims to demonstrate the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features while fully complying with the Generation IV goal of minor actinide (MA) burning capability. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Simplicity is expected to reduce both the capital cost and the construction time; these are also supported by the compactness of the reactor building (reduced footprint and height). The reduced footprint would be possible due to the elimination of the Intermediate Cooling System, the reduced elevation the result of the design approach of reduced-height components.« less

  18. The effects of stainless steel radial reflector on core reactivity for small modular reactor

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

    Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr; Cho, Sung Ju, E-mail: sungju@knfc.co.kr

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5more » w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.« less

  19. Hanford Laboratories Operation monthly activities report, September 1961

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

    Not Available

    1961-10-16

    This is the monthly report for the Hanford Laboratories Operation September 1961. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

  20. Radiological Protection Issues Arising During and After the Fukushima Nuclear Reactor Accident-Memorandum of TG 84 of ICRP.

    PubMed

    Weiss, Wolfgang

    2016-09-01

    Observations and lessons identified after the Fukushima accident have been collected and assessed by ICRP Task Group 84. Together with the observations of other expert organizations, they are being used to further develop the current system of protection. While many of the established protection criteria remain valid, improvements are needed in three areas. Key issues related to the need of planning for long-term protective actions (criteria for returning home, dealing with waste) have to be implemented as important elements of the national protection strategies during the preparedness stage. The justification of disruptive protective actions and the protection of vulnerably groups of the population need to be reconsidered to avoid unpleasant imbalances and outcomes. The coexistence of radiation-induced health effects and health effects with social determinants requires consideration of both aspects in decision-making and response. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  1. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

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

    Moses, David Lewis

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physicalmore » Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed commercially and have only been demonstrated in testing at a laboratory scale.« less

  2. Superphenix: Restarting, with emphasis on research

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

    NONE

    1994-04-01

    French Prime Minister Edouard Baladour announced on February 22 that the Superphenix fast reactor at Creys-Malville will be allowed to restart. Along with his industry minister, Gerard Longuet, and the environment minister, Michel Bamier, Baladour has accepted the recommendation for reissuance of an operating license submitted to the ministers in January by the nuclear installations safety directorate (DSIN). In announcing their approval for restart, the ministers emphasized that the reactor was to be used as a research and demonstration facility rather than a power station. In particular, they said it should be used to investigate the possibility of using fastmore » reactors as plutonium burners rather than breeders and for the incineration of other long-lived actinide wastes. This point was made also by DSIN Director Andre-Claude Lacoste in his recommendations, but appears to have been given more weight by the government ministers. The ministers also accepted the DSIN recommendation that restart be conditional on completion of improvements to fire protection systems for the secondary sodium coolant ducts. Baladour said that it would take several months to complete this work, but other sources have suggested that the reactor could be ready to operate this month. DSIN has also stipulated that the power level be limited for several months to around 50 percent of the plant`s 1200-MWe potential in order to check out all the systems carefully.« less

  3. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

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

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

  4. The ENEA criticality accident dosimetry system: a contribution to the 2002 international intercomparison at the SILENE reactor.

    PubMed

    Gualdrini, G; Bedogni, R; Fantuzzi, E; Mariotti, F

    2004-01-01

    The present paper summarises the activity carried out at the ENEA Radiation Protection Institute for updating the methodologies employed for the evaluation of the neutron and photon dose to the exposed workers in case of a criticality accident, in the framework of the 'International Intercomparison of Criticality Accident Dosimetry Systems' (Silène reactor, IRSN-CEA-Valduc June 2002). The evaluation of the neutron spectra and the neutron dosimetric quantities relies on activation detectors and on unfolding algorithms. Thermoluminescent detectors are employed for the gamma dose measurement. The work is aimed at accurately characterising the measurement system and, at the same time, testing the algorithms. Useful spectral information were included, based on Monte Carlo simulations, to take into account the potential accident scenarios of practical interest. All along this exercise intercomparison a particular attention was devoted to the 'traceability' of all the experimental and computational parameters and therefore, aimed at an easy treatment by the user.

  5. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    NASA Astrophysics Data System (ADS)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  6. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    NASA Astrophysics Data System (ADS)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  7. Enhancement of NRC station blackout requirements for nuclear power plants

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

    McConnell, M. W.

    2012-07-01

    The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50,more » Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)« less

  8. SPERTI Reactor Pit Building (PER605). Earth shielding protect adjacent Instrument ...

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

    SPERT-I Reactor Pit Building (PER-605). Earth shielding protect adjacent Instrument Cell (PER-606). Security fencing surrounds complex, to which gate entry is provided next to Guard House (PER-607). Note gravel road leading to control area. Earth-covered conduit leads from instrument cell to terminal building out of view. Photographer: R.G. Larsen. Date: June 22, 1955. INEEL negative no. 55-1701 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  9. 10 CFR 51.121 - Status of NEPA actions.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

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

  10. 10 CFR 51.121 - Status of NEPA actions.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

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

  11. 10 CFR 51.121 - Status of NEPA actions.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

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

  12. 10 CFR 72.180 - Physical protection plan.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

  13. 10 CFR 72.180 - Physical protection plan.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

  14. 10 CFR 72.180 - Physical protection plan.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

  15. 10 CFR 72.180 - Physical protection plan.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

  16. 10 CFR 72.180 - Physical protection plan.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

  17. Corium protection assembly

    DOEpatents

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A corium protection assembly includes a perforated base grid disposed below a pressure vessel containing a nuclear reactor core and spaced vertically above a containment vessel floor to define a sump therebetween. A plurality of layers of protective blocks are disposed on the grid for protecting the containment vessel floor from the corium.

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

    Palomar, J.; Wyman, R.

    This document provides recommendations to guide reviewers in the application of Programmable Logic Controllers (PLCS) to the control, monitoring and protection of nuclear reactors. The first topics addressed are system-level design issues, specifically including safety. The document then discusses concerns about the PLC manufacturing organization and the protection system engineering organization. Supplementing this document are two appendices. Appendix A summarizes PLC characteristics. Specifically addressed are those characteristics that make the PLC more suitable for emergency shutdown systems than other electrical/electronic-based systems, as well as characteristics that improve reliability of a system. Also covered are PLC characteristics that may create anmore » unsafe operating environment. Appendix B provides an overview of the use of programmable logic controllers in emergency shutdown systems. The intent is to familiarize the reader with the design, development, test, and maintenance phases of applying a PLC to an ESD system. Each phase is described in detail and information pertinent to the application of a PLC is pointed out.« less

  19. Reactor operation environmental information document

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

    Haselow, J.S.; Price, V.; Stephenson, D.E.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimalmore » impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.« less

  20. Education and training in the field of nuclear instrumentation and measurement: CEA/INSTN (National Institute for Nuclear Sciences and Technologies) strategy to improve and develop new pedagogical tools and methods

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

    Vitart, Xavier; Foulon, Francois; Bodineau, Jean Christophe

    Part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) is a higher education institution whose mission is to provide students and professionals a high level of scientific and technological qualification in all disciplines related to nuclear energy applications. In this frame, INSTN carries out education and training (E and T) programs in nuclear instrumentation and radioprotection. Its strategy has always been to complete theoretical courses by training courses and laboratory works carried out on an extensive range of training tools that includes a large panel of nuclear instrumentation asmore » well as software applications. Since its creation in 1956, the INSTN has conducted both education and vocational programs on ionizing radiation detection. An extensive range of techniques have commonly been used during practical works with students and employees of companies who need to get the knowledge and specialization in this field. Today, the INSTN is mainly equipped with usual detectors and electronics in large numbers in order to be able to accommodate up to 48 trainees at the same time in two classrooms, with only two trainees for one workstation in order to optimize their learning. In the field of the neutron detection systems, the INSTN has strongly developed its offer taking advantage of the use of research reactors, such as ISIS reactor (700 kW) at Saclay. The implementation of neutron detection systems specific to the courses offers a unique way of observing and analysing the signal coming from neutron detectors, as well as learning how to set the parameters of the detection system in real conditions. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. Another field of activity of the INSTN is the development of new teaching tools using software applications. We describe hereafter two applications that have recently been developed by the institute, O.S.I.R.I.S and DOSIMEX. The INSTN and the company OREKA have developed an innovative teaching tool named O.S.I.R.I.S. The tool is built on a virtual 3D environment in which users operate in a totally free way. The action is located in a pressurised water reactor in the environment of a steam generator building. The users are students or professionals who want to learn and train on concrete situations on how to protect the workers against radiation. Through the use of this serious game, users must: establish a predictive dose evaluation, implement the principles of radiation protection, perform the dose result of the operation, analyse the gap between the predictive collective dose and the collective dose achieved, and decide about the lessons to provide feedback. Taking into account the increasing constraints related to radiation protection, it is essential to use computer codes to evaluate the dose rates generated by sources of ionizing radiation. Many powerful codes exist but are often relatively complex to implement and are dedicated to experts. In addition, these codes are often 'black boxes' that does not allow the user to understand the underlying physics. The objective of DOSIMEX codes developed by the INSTN is to give, by the means of a VBA code with a simple interface, the ability to easily conduct a wide range of radiological exposure situations: calculation of gamma, beta, neutron equivalent dose rate, calculation of internal contamination and atmospheric transfer.« less

  1. Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

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

    Tiberi, V.

    2012-07-01

    The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

  2. Process for direct conversion of reactive metals to glass

    DOEpatents

    Rajan, John B.; Kumar, Romesh; Vissers, Donald R.

    1990-01-01

    Radioactive alkali metal is introduced into a cyclone reactor in droplet form by an aspirating gas. In the cyclone metal reactor the aspirated alkali metal is contacted with silica powder introduced in an air stream to form in one step a glass. The sides of the cyclone reactor are preheated to ensure that the initial glass formed coats the side of the reactor forming a protective coating against the reactants which are maintained in excess of 1000.degree. C. to ensure the formation of glass in a single step.

  3. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

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

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilisticmore » risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.« less

  4. Radiological protection issues arising during and after the Fukushima nuclear reactor accident.

    PubMed

    González, Abel J; Akashi, Makoto; Boice, John D; Chino, Masamichi; Homma, Toshimitsu; Ishigure, Nobuhito; Kai, Michiaki; Kusumi, Shizuyo; Lee, Jai-Ki; Menzel, Hans-Georg; Niwa, Ohtsura; Sakai, Kazuo; Weiss, Wolfgang; Yamashita, Shunichi; Yonekura, Yoshiharu

    2013-09-01

    Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with 'contamination' of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of potential health effects are properly interpreted; the limitations of epidemiological studies for attributing radiation effects following low exposures are understood; any confusion on protection quantities and units is resolved; the potential hazard from the intake of radionuclides into the body is elucidated; rescuers and volunteers are protected with an ad hoc system; clear recommendations on crisis management and medical care and on recovery and rehabilitation are available; recommendations on public protection levels (including infant, children and pregnant women and their expected offspring) and associated issues are consistent and understandable; updated recommendations on public monitoring policy are available; acceptable (or tolerable) 'contamination' levels are clearly stated and defined; strategies for mitigating the serious psychological consequences arising from radiological accidents are sought; and, last but not least, failures in fostering information sharing on radiological protection policy after an accident need to be addressed with recommendations to minimise such lapses in communication.

  5. OAST Space Theme Workshop. Volume 2: Theme summary. 4: Solar system exploration (no. 10). A: Statement of theme: B. 26 April 1976 Presentation. C. Summary. D. Initiative actions (form 5)

    NASA Technical Reports Server (NTRS)

    1976-01-01

    Major strategies for exploring the solar system focus on the return of information and the return of matter. Both the planetary exploration facility, and an orbiting automated space station, and the sample return and exploration facility have similar requirements. The single most essential need to enable intensive study of the outer solar system is nuclear propulsion and power capability. New initiatives in 1978 related to the reactor, data and sample acquisition and return, navigation, and environmental protection are examined.

  6. Pulsed corona generation using a diode-based pulsed power generator

    NASA Astrophysics Data System (ADS)

    Pemen, A. J. M.; Grekhov, I. V.; van Heesch, E. J. M.; Yan, K.; Nair, S. A.; Korotkov, S. V.

    2003-10-01

    Pulsed plasma techniques serve a wide range of unconventional processes, such as gas and water processing, hydrogen production, and nanotechnology. Extending research on promising applications, such as pulsed corona processing, depends to a great extent on the availability of reliable, efficient and repetitive high-voltage pulsed power technology. Heavy-duty opening switches are the most critical components in high-voltage pulsed power systems with inductive energy storage. At the Ioffe Institute, an unconventional switching mechanism has been found, based on the fast recovery process in a diode. This article discusses the application of such a "drift-step-recovery-diode" for pulsed corona plasma generation. The principle of the diode-based nanosecond high-voltage generator will be discussed. The generator will be coupled to a corona reactor via a transmission-line transformer. The advantages of this concept, such as easy voltage transformation, load matching, switch protection and easy coupling with a dc bias voltage, will be discussed. The developed circuit is tested at both a resistive load and various corona reactors. Methods to optimize the energy transfer to a corona reactor have been evaluated. The impedance matching between the pulse generator and corona reactor can be significantly improved by using a dc bias voltage. At good matching, the corona energy increases and less energy reflects back to the generator. Matching can also be slightly improved by increasing the temperature in the corona reactor. More effective is to reduce the reactor pressure.

  7. AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

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

    Pfister, A.; Goossen, C.; Coogler, K.

    2012-07-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plantmore » is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is further surrounded by a substantial 'steel concrete' composite shield building. The containment vessel is not affected by external flooding, and the shield building design provides hazard protection beyond that provided by a comparable reinforced concrete structure. The intent of this paper is to demonstrate the robustness of the AP1000 design against extreme events. The paper will focus on the plants ability to withstand extreme external events such as beyond design basis flooding, seismic events, and malicious aircraft impact. The paper will highlight the robustness of the AP1000 nuclear island design including the protection provided by the unique AP1000 composite shield building. (authors)« less

  8. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  9. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  10. 76 FR 1376 - Physical Protection of Irradiated Reactor Fuel in Transit Extension of Comment Period

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-10

    ... dockets to Carol Gallagher, telephone: 301-492- 3668, e-mail: [email protected] . Mail comments to... Adjudications Staff. E-mail comments to: [email protected] . If you do not receive a reply e-mail..., e-mail: [email protected] . SUPPLEMENTARY INFORMATION: Submitting Comments and Accessing...

  11. Digital Signal Processing Methods for Safety Systems Employed in Nuclear Power Industry

    NASA Astrophysics Data System (ADS)

    Popescu, George

    Some of the major safety concerns in the nuclear power industry focus on the readiness of nuclear power plant safety systems to respond to an abnormal event, the security of special nuclear materials in used nuclear fuels, and the need for physical security to protect personnel and reactor safety systems from an act of terror. Routine maintenance and tests of all nuclear reactor safety systems are performed on a regular basis to confirm the ability of these systems to operate as expected. However, these tests do not determine the reliability of these safety systems and whether the systems will perform for the duration of an accident and whether they will perform their tasks without failure after being engaged. This research has investigated the progression of spindle asynchronous error motion determined from spindle accelerations to predict bearings failure onset. This method could be applied to coolant pumps that are essential components of emergency core cooling systems at all nuclear power plants. Recent security upgrades mandated by the Nuclear Regulatory Commission and the Department of Homeland Security have resulted in implementation of multiple physical security barriers around all of the commercial and research nuclear reactors in the United States. A second part of this research attempts to address an increased concern about illegal trafficking of Special Nuclear Materials (SNM). This research describes a multi element scintillation detector system designed for non - invasive (passive) gamma ray surveillance for concealed SNM that may be within an area or sealed in a package, vehicle or shipping container. Detection capabilities of the system were greatly enhanced through digital signal processing, which allows the combination of two very powerful techniques: 1) Compton Suppression (CS) and 2) Pulse Shape Discrimination (PSD) with less reliance on complicated analog instrumentation.

  12. Current status of radiological protection at nuclear power stations in Japan.

    PubMed

    Suzuki, Akira; Hori, Shunsuke

    2011-07-01

    The radiation dose to workers at nuclear power stations (NPSs) in Japan was drastically reduced between the late-1970s and the early-1990s by continuous dose-reduction programmes. The total collective dose of radiation workers in FY 2008 was 84.04 person Sv, while the average collective dose was 1.5 person Sv per reactor. The average annual individual dose was 1.1 mSv and the maximum annual individual dose was 19.5 mSv. These values are sufficiently lower than the regulatory dose limits. Radioactive effluent released from NPSs is already so trivial that additional protective measures will not be necessary. Experience in radiation protection at NPSs has been accumulated over 40 y and will be very useful in establishing a rational radiation control system in the future.

  13. Pulsed Corona Discharge Generated By Marx Generator

    NASA Astrophysics Data System (ADS)

    Sretenovic, G. B.; Obradovic, B. M.; Kovacevic, V. V.; Kuraica, M. M.; Puric J.

    2010-07-01

    The pulsed plasma has a significant role in new environmental protection technologies. As a part of a pulsed corona system for pollution control applications, Marx type repetitive pulse generator was constructed and tested in arrangement with wire-plate corona reactor. We performed electrical measurements, and obtained voltage and current signals, and also power and energy delivered per pulse. Ozone formation by streamer plasma in air was chosen to monitor chemical activity of the pulsed corona discharge.

  14. Recent Advances in Pd-Based Membranes for Membrane Reactors.

    PubMed

    Arratibel Plazaola, Alba; Pacheco Tanaka, David Alfredo; Van Sint Annaland, Martin; Gallucci, Fausto

    2017-01-01

    Palladium-based membranes for hydrogen separation have been studied by several research groups during the last 40 years. Much effort has been dedicated to improving the hydrogen flux of these membranes employing different alloys, supports, deposition/production techniques, etc. High flux and cheap membranes, yet stable at different operating conditions are required for their exploitation at industrial scale. The integration of membranes in multifunctional reactors (membrane reactors) poses additional demands on the membranes as interactions at different levels between the catalyst and the membrane surface can occur. Particularly, when employing the membranes in fluidized bed reactors, the selective layer should be resistant to or protected against erosion. In this review we will also describe a novel kind of membranes, the pore-filled type membranes prepared by Pacheco Tanaka and coworkers that represent a possible solution to integrate thin selective membranes into membrane reactors while protecting the selective layer. This work is focused on recent advances on metallic supports, materials used as an intermetallic diffusion layer when metallic supports are used and the most recent advances on Pd-based composite membranes. Particular attention is paid to improvements on sulfur resistance of Pd based membranes, resistance to hydrogen embrittlement and stability at high temperature.

  15. Westinghouse Small Modular Reactor passive safety system response to postulated events

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

    Smith, M. C.; Wright, R. F.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. Themore » integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)« less

  16. An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

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

    Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

    2010-01-01

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

  17. 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.

  18. Safety in the Chemical Laboratory: Handling of Oxygen in Research Experiments.

    ERIC Educational Resources Information Center

    Burnett, R. J.; Cole, J. E., Jr.

    1985-01-01

    Examines some of the considerations involved in setting up a typical oxygen/organic reaction. These considerations (including protection for personnel/equipment, adequate ventilation, reactor design, maximum reactor charge, operating procedures, and others) influence how the reaction is to be conducted and what compromises the scientist must…

  19. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    NASA Astrophysics Data System (ADS)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  20. Hardwired Control Changes For NSTX DC Power Feeds

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

    Ramakrishnan, S.

    The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The original TFTR Hardwired Control System (HCS) with electromechanical relays was used for NSTX DC Power loop control and protection during NSTX operations. As part of the NSTX Upgrade, the HCS is being changed to a PLC-based system with the same control logic. This paper gives a description ofmore » the changeover to the new PLC-based system __________________________________________________« less

  1. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... shall: (i) Design, construct, install and maintain physical barriers as necessary to control access into.... (10) Vehicle control measures. Consistent with the physical protection program design requirements of... maintain vehicle control measures, as necessary, to protect against the design basis threat of radiological...

  2. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... shall: (i) Design, construct, install and maintain physical barriers as necessary to control access into.... (10) Vehicle control measures. Consistent with the physical protection program design requirements of... maintain vehicle control measures, as necessary, to protect against the design basis threat of radiological...

  3. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... shall: (i) Design, construct, install and maintain physical barriers as necessary to control access into.... (10) Vehicle control measures. Consistent with the physical protection program design requirements of... maintain vehicle control measures, as necessary, to protect against the design basis threat of radiological...

  4. Protective interior wall and attaching means for a fusion reactor vacuum vessel

    DOEpatents

    Phelps, R.D.; Upham, G.A.; Anderson, P.M.

    1985-03-01

    The wall basically consists of an array of small rectangular plates attached to the existing walls with threaded fasteners. The protective wall effectively conceals and protects all mounting hardware beneath the plate array, while providing a substantial surface area that will absorb plasma energy.

  5. Lessons Learned in Protection of the Public for the Accident at the Fukushima Daiichi Nuclear Power Plant.

    PubMed

    Callen, Jessica; Homma, Toshimitsu

    2017-06-01

    What insights can the accident at the Fukushima Daiichi nuclear power plant provide in the reality of decision making on actions to protect the public during a severe reactor and spent fuel pool emergency? In order to answer this question, and with the goal of limiting the consequences of any future emergencies at a nuclear power plant due to severe conditions, this paper presents the main actions taken in response to the emergency in the form of a timeline. The focus of this paper is those insights concerning the progression of an accident due to severe conditions at a light water reactor nuclear power plant that must be understood in order to protect the public.

  6. Potential effects of the fire protection system sprays at Browns Ferry on fission product transport

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

    Niemczyk, S.J.

    1983-01-01

    The fire protection system (FPS) sprays within any nuclear plant are not intended to mitigate radioactive releases to the environment resulting from severe core-damage accidents. However, it has been shown here that during certain postulated severe accident scenarios at the Browns Ferry Nuclear Plant, the functioning of FPS sprays could have a significant impact on the radioactive releases. Thus the effects of those sprays need to be taken into account for realistic estimation of source terms for some accident scenarios. The effects would include direct ones such as cooling of the reactor building atmosphere and scrubbing of radioactivity from it,more » as well as indirect effects such as an altered likelihood of hydrogen burning and flooding of various safety-related pumps in the reactor building basement. Thus some of the impacts of the sprays would be beneficial with respect to mitigating releases to the environment but some others might not be. The effects of the FPS would be very scenario dependent with a wide range of potential effects often existing for a given accident sequence. Any generalization of the specific results presented here for Browns Ferry to other nuclear plants must be done cautiously, as it appears from a preliminary investigation that the relevant physical and operational characteristics of FPS spray systems differ widely among even otherwise apparently similar plants. Likewise the standby gas treatment systems, which substantially impact the effects of the FPS, differ significantly among plants. More work for both Mark I plants and other plants, BWRs and PWRs alike, is indicated so the potential effects of FPS spray systems during severe accidents can be at least ball-parked for more realistic accident analyses.« less

  7. 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. .

  8. 61. VIEW LOOKING NORTHWEST AT A SIGNAL REACTOR OR CHOKE ...

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

    61. VIEW LOOKING NORTHWEST AT A SIGNAL REACTOR OR CHOKE COIL. WITHIN THE PROTECTIVE ENCLOSURE IS AN AIR AND PORCELAIN INSULATED COIL OF 5/8' DIAMETER STRANDED COPPER WIRE. REACTOR COILS WERE PLACED IN SERIES WITH EACH LEG OF THREE PHASE GENERATORS. THEIR FUNCTION WAS TO MODERATE SURGES OF CURRENT CAUSED BY LIGHTNING STRIKES, OPEN OR SHORT CIRCUIT PROBLEMS ON THE LINE. - New York, New Haven & Hartford Railroad, Cos Cob Power Plant, Sound Shore Drive, Greenwich, Fairfield County, CT

  9. 10 CFR 51.23 - Temporary storage of spent fuel after cessation of reactor operation-generic determination of no...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Temporary storage of spent fuel after cessation of reactor operation-generic determination of no significant environmental impact. 51.23 Section 51.23 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED...

  10. 75 FR 35001 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-21

    ... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee AGENCY: Department of Energy, Office of Nuclear Energy. ACTION: Notice of Open... facsimile (202) 586- 0544; e-mail [email protected]nuclear.energy.gov . Additional information may also be...

  11. 75 FR 61139 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-04

    ... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee AGENCY: Department of Energy, Office of Nuclear Energy. ACTION: Notice of Open...) 586- 0544; e-mail [email protected]nuclear.energy.gov . Additional information will be available at http...

  12. Nuclear design of a very-low-activation fusion reactor

    NASA Astrophysics Data System (ADS)

    Cheng, E. T.; Hopkins, G. R.

    1983-06-01

    The nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications were investigated. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a Tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE Tokamak reactor design.

  13. Nuclear Technology. Course 30: Mechanical Inspection. Module 30-6, Protective Coating Inspection.

    ERIC Educational Resources Information Center

    Espy, John

    This sixth in a series of eight modules for a course titled Mechanical Inspection describes the duties of the nuclear quality assurance/quality control technician that are associated with protective coatings, and the national standards that govern the selection, application, and inspection of protective coatings for the reactor containment…

  14. 76 FR 9379 - Exelon Generation Company, LLC; Lasalle County Station, Units 1 and 2; Environmental Assessment...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-02-17

    ...-38, ``Storage of Low-Level Radioactive Wastes at Power Reactor Sites'' and to meet the radiation protection standards in 10 CFR Part 20, ``Standards for Protection Against Radiation,'' and 40 CFR Part 190, ``Environmental Radiation Protection Standards for Nuclear Power Operations.'' Environmental Impacts of the...

  15. [Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

    PubMed

    Kim, D S

    2012-01-01

    The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

  16. Start-up of a multi-stage system for biogas production and solid waste treatment in low-tech countries.

    PubMed

    Biey, E M; Musibono, E D; Verstraete, W

    2003-01-01

    Vegetable fruit garden wastes were treated anaerobically using a multistage Dranco system. The digesters were composed of three 50 L vessels kept in mesophilic conditions. They were operating at 14.5-17% TS. By controlling the pH in the system, the start-up for biogas production was shortened to 60 days. The pH correction was a buffering which enhanced methanogenic activity in the digesters. With a loading rate of 4.1 kg VS/m3 reactor/day, the production of biogas was 5 m3/m3 reactor/ day, and 60-70% methane content. This allowed making a multisystem by starting every 3 weeks with new vessels in order to maintain biogas production, to be used in industries or in local communities in low-tech countries. The designed model was started in Kinshasa (Congo) where a project is expected to treat one ton of solid waste on a daily basis, for a production of 100 m3 biogas. This cost effectiveness of the system is demonstrated and presents the opportunity for biowaste treatment coupled with environmental protection and substantial energy recovery.

  17. Fast Time Response Electromagnetic Disruption Mitigation Concept

    DOE PAGES

    Raman, R.; Jarboe, T.; Jernigan, Thomas C.; ...

    2015-09-28

    An important and urgent issue for ITER is predicting and controlling disruptions. Tokamaks and spherical tokamaks have the potential to disrupt. Methods to rapidly quench the discharge after an impending disruption is detected are essential to protect the vessel and internal components. The warning time for the onset of some disruptions in tokamaks could be <10 ms, which poses stringent requirements on the disruption mitigation system for reactor systems. In this proposed method, a cylindrical boron nitride projectile containing a radiative payload composed of boron, boron nitride, or beryllium particulate matter and weighing similar to 15 g is accelerated tomore » velocities on the order of 1 to 2 km/s in <2 ms in a linear rail gun accelerator. A partially fragmented capsule is then injected into the tokamak discharge in the 3- to 6-ms timescale, where the radiative payload is dispersed. The device referred to as an electromagnetic particle injector has the potential to meet the short warning timescales for which a reactor disruption mitigation system must be built. The system is fully electromagnetic, with no mechanical moving parts, which ensures high reliability after a period of long standby.« less

  18. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

    Su'ud, Zaki; Anshari, Rio

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

  19. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Anshari, Rio

    2012-06-01

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  20. 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.

  1. Experimental Validation of a Resilient Monitoring and Control System

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

    Wen-Chiao Lin; Kris R. E. Villez; Humberto E. Garcia

    2014-05-01

    Complex, high performance, engineering systems have to be closely monitored and controlled to ensure safe operation and protect public from potential hazards. One of the main challenges in designing monitoring and control algorithms for these systems is that sensors and actuators may be malfunctioning due to malicious or natural causes. To address this challenge, this paper addresses a resilient monitoring and control (ReMAC) system by expanding previously developed resilient condition assessment monitoring systems and Kalman filter-based diagnostic methods and integrating them with a supervisory controller developed here. While the monitoring and diagnostic algorithms assess plant cyber and physical health conditions,more » the supervisory controller selects, from a set of candidates, the best controller based on the current plant health assessments. To experimentally demonstrate its enhanced performance, the developed ReMAC system is then used for monitoring and control of a chemical reactor with a water cooling system in a hardware-in-the-loop setting, where the reactor is computer simulated and the water cooling system is implemented by a machine condition monitoring testbed at Idaho National Laboratory. Results show that the ReMAC system is able to make correct plant health assessments despite sensor malfunctioning due to cyber attacks and make decisions that achieve best control actions despite possible actuator malfunctioning. Monitoring challenges caused by mismatches between assumed system component models and actual measurements are also identified for future work.« less

  2. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  3. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  4. Continuously-stirred anaerobic digester to convert organic wastes into biogas: system setup and basic operation.

    PubMed

    Usack, Joseph G; Spirito, Catherine M; Angenent, Largus T

    2012-07-13

    Anaerobic digestion (AD) is a bioprocess that is commonly used to convert complex organic wastes into a useful biogas with methane as the energy carrier. Increasingly, AD is being used in industrial, agricultural, and municipal waste(water) treatment applications. The use of AD technology allows plant operators to reduce waste disposal costs and offset energy utility expenses. In addition to treating organic wastes, energy crops are being converted into the energy carrier methane. As the application of AD technology broadens for the treatment of new substrates and co-substrate mixtures, so does the demand for a reliable testing methodology at the pilot- and laboratory-scale. Anaerobic digestion systems have a variety of configurations, including the continuously stirred tank reactor (CSTR), plug flow (PF), and anaerobic sequencing batch reactor (ASBR) configurations. The CSTR is frequently used in research due to its simplicity in design and operation, but also for its advantages in experimentation. Compared to other configurations, the CSTR provides greater uniformity of system parameters, such as temperature, mixing, chemical concentration, and substrate concentration. Ultimately, when designing a full-scale reactor, the optimum reactor configuration will depend on the character of a given substrate among many other nontechnical considerations. However, all configurations share fundamental design features and operating parameters that render the CSTR appropriate for most preliminary assessments. If researchers and engineers use an influent stream with relatively high concentrations of solids, then lab-scale bioreactor configurations cannot be fed continuously due to plugging problems of lab-scale pumps with solids or settling of solids in tubing. For that scenario with continuous mixing requirements, lab-scale bioreactors are fed periodically and we refer to such configurations as continuously stirred anaerobic digesters (CSADs). This article presents a general methodology for constructing, inoculating, operating, and monitoring a CSAD system for the purpose of testing the suitability of a given organic substrate for long-term anaerobic digestion. The construction section of this article will cover building the lab-scale reactor system. The inoculation section will explain how to create an anaerobic environment suitable for seeding with an active methanogenic inoculum. The operating section will cover operation, maintenance, and troubleshooting. The monitoring section will introduce testing protocols using standard analyses. The use of these measures is necessary for reliable experimental assessments of substrate suitability for AD. This protocol should provide greater protection against a common mistake made in AD studies, which is to conclude that reactor failure was caused by the substrate in use, when really it was improper user operation.

  5. PBF Reactor Building (PER620). Cubicle 10 area in basement. Highdensity ...

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

    PBF Reactor Building (PER-620). Cubicle 10 area in basement. High-density shielding bricks will protect personnel from radiation coming from in-pile-tube coolant and blowdown tank. Photographer: John Capek. Date: January 26, 1970. INEEL negative no. 70-348 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  6. 77 FR 27487 - License Amendment Request From The State University of New York, University of Buffalo Reactor...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-10

    ... State University of New York, University of Buffalo Reactor Facility AGENCY: Nuclear Regulatory... staff at 1-800-397-4209, 301-415- 4737, or by email to [email protected] . The University of Buffalo... license amendment application from the State University of New York, University of Buffalo requesting...

  7. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  8. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  9. Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

    NASA Technical Reports Server (NTRS)

    Sibille, Laurent; Dominques, Jesus A.

    2012-01-01

    The maturation of Molten Regolith Electrolysis (MRE) as a viable technology for oxygen and metals production on explored planets relies on the realization of the self-heating mode for the reactor. Joule heat generated during regolith electrolysis creates thermal energy that should be able to maintain the molten phase (similar to electrolytic Hall-Heroult process for aluminum production). Self-heating via Joule heating offers many advantages: (1) The regolith itself is the crucible material, it protects the vessel walls (2) Simplifies the engineering of the reactor (3) Reduces power consumption (no external heating) (4) Extends the longevity of the reactor. Predictive modeling is a tool chosen to perform dimensional analysis of a self-heating reactor: (1) Multiphysics modeling (COMSOL) was selected for Joule heat generation and heat transfer (2) Objective is to identify critical dimensions for first reactor prototype.

  10. System and method for air temperature control in an oxygen transport membrane based reactor

    DOEpatents

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  11. System and method for temperature control in an oxygen transport membrane based reactor

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

    Kelly, Sean M.

    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  12. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  13. Development of toroid-type HTS DC reactor series for HVDC system

    NASA Astrophysics Data System (ADS)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-11-01

    This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  14. The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

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

    None, None

    NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise requiredmore » to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.« less

  15. 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.

  16. 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

  17. The pre-conceptual design of the nuclear island of ASTRID

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

    Saez, M.; Menou, S.; Uzu, B.

    The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main componentsmore » with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)« less

  18. Innovative Use of Cr(VI) Plume Depictions and Pump-and-Treat Capture Analysis to Estimate Risks of Contaminant Discharge to Surface Water at Hanford Reactor Areas

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

    Miller, Chuck W.; Hanson, James P.; Ivarson, Kristine A.

    2015-01-14

    The Hanford Site nuclear reactor operations required large quantities of high-quality cooling water, which was treated with chemicals including sodium dichromate dihydrate for corrosion control. Cooling water leakage, as well as intentional discharge of cooling water to ground during upset conditions, produced extensive groundwater recharge mounds consisting largely of contaminated cooling water and resulted in wide distribution of hexavalent chromium (Cr[VI]) contamination in the unconfined aquifer. The 2013 Cr(VI) groundwater plumes in the 100 Areas cover approximately 6 km2 (1500 acres), primarily in the 100-HR-3 and 100-KR-4 groundwater operable units (OUs). The Columbia River is a groundwater discharge boundary; wheremore » the plumes are adjacent to the Columbia River there remains a potential to discharge Cr(VI) to the river at concentrations above water quality criteria. The pump-and-treat systems along the River Corridor are operating with two main goals: 1) protection of the Columbia River, and 2) recovery of contaminant mass. An evaluation of the effectiveness of the pump-and-treat systems was needed to determine if the Columbia River was protected from contamination, and also to determine where additional system modifications may be needed. In response to this need, a technique for assessing the river protection was developed which takes into consideration seasonal migration of the plume and hydraulic performance of the operating well fields. Groundwater contaminant plume maps are generated across the Hanford Site on an annual basis. The assessment technique overlays the annual plume and the capture efficiency maps for the various pump and treat systems. The river protection analysis technique was prepared for use at the Hanford site and is described in detail in M.J. Tonkin, 2013. Interpolated capture frequency maps, based on mapping dynamic water level observed in observation wells and derived water levels in the vicinity of extraction and injection wells, are developed initially. Second, simulated capture frequency maps are developed, based on transport modelling results. Both interpolated and simulated capture frequency maps are based on operation of the systems over a full year. These two capture maps are then overlaid on the plume distribution maps for inspection of the relative orientation of the contaminant plumes with the capture frequency. To quantify the relative degree of protection of the river from discharges of Cr(VI) (and conversely, the degree of threat) at any particular location, a systematic method of evaluating and mapping the plume/capture relationship was developed. By comparing the spatial relationship between contaminant plumes and hydraulic capture frequency, an index of relative protectiveness is developed and the results posted on the combined plume/capture plan view map. Areas exhibiting lesser degrees of river protection are identified for remedial process optimization actions to control plumes and prevent continuing discharge of Cr(VI) to the river.« less

  19. Upper internals arrangement for a pressurized water reactor

    DOEpatents

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  20. Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

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

    Ambrose, T.W.

    1965-06-04

    Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

  1. Integral reactor system and method for fuel cells

    DOEpatents

    Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.

    2017-03-07

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  2. Integral reactor system and method for fuel cells

    DOEpatents

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  3. Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

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

    Moir, R. W.

    2012-06-19

    Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroymore » fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.« less

  4. Future energy system in environment, economy, and energy problems (2) various nuclear energy system evaluations

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

    Matsui, Kazuaki; Ujita, Hiroshi; Tashimo, Masanori

    2006-07-01

    Role and potentials of nuclear energy system in the energy options are discussed from the viewpoint of sustainable development with protecting from global warming by using the energy module structure of GRAPE model. They change and are affected dramatically by different sets of energy characteristics, nuclear behavior and energy policy even under the moderate set of presumptions. Introduction of thousands of reactors in the end of the century seems inevitable for better life and cleaner earth, but it will not come without efforts and cost. The analysis suggests the need of long term planning and R and D efforts undermore » the wisdom. (authors)« less

  5. Magnetic Materials Suitable for Fission Power Conversion in Space Missions

    NASA Technical Reports Server (NTRS)

    Bowman, Cheryl L.

    2012-01-01

    Terrestrial fission reactors use combinations of shielding and distance to protect power conversion components from elevated temperature and radiation. Space mission systems are necessarily compact and must minimize shielding and distance to enhance system level efficiencies. Technology development efforts to support fission power generation scenarios for future space missions include studying the radiation tolerance of component materials. The fundamental principles of material magnetism are reviewed and used to interpret existing material radiation effects data for expected fission power conversion components for target space missions. Suitable materials for the Fission Power System (FPS) Project are available and guidelines are presented for bounding the elevated temperature/radiation tolerance envelope for candidate magnetic materials.

  6. THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

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

    Colley, J.R.

    1962-12-01

    The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

  7. Thermochemical reactor systems and methods

    DOEpatents

    Lipinski, Wojciech; Davidson, Jane Holloway; Chase, Thomas Richard

    2016-11-29

    Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

  8. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  9. PBF Reactor Building (PER620). Camera is facing east and down ...

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

    PBF Reactor Building (PER-620). Camera is facing east and down into canal and storage pit for fuel rod assemblies. Stainless steel liner is being applied, temporarily covered with plywood for protection. Photographer: John Capek. Date: August 29, 1969. INEEL negative no. 69-4641 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  10. A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

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

    Petti, David Andrew

    2017-04-01

    Modular high temperature gas-cooled reactor (HTGR) designs were developed to provide natural safety, which prevents core damage under all licensing basis events. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. The required level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude relative to source terms for other reactor types and allows a graded approach to emergency planning and the potential elimination of the need for evacuation and sheltering beyond a small exclusion area. Achieving this level, however,more » is predicated on exceptionally high coated-particle fuel fabrication quality and excellent performance under normal operation and accident conditions. The design goal of modular HTGRs is to meet the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) for offsite dose at the Exclusion Area Boundary (EAB). To achieve this, the reactor design concepts require a level of fuel integrity that is far better than that achieved for all prior U.S.-manufactured tristructural isotropic (TRISO) coated particle fuel.« less

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

    Kiselyov, V.A.; Sokov, L.M.

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it hasmore » been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.« less

  12. Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

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

    Smith, Curtis; Rabiti, Cristian; Martineau, Richard

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degreemore » of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”« less

  13. Space nuclear power systems 1989; Proceedings of the 6th Symposium, Albuquerque, NM, Jan. 8-12, 1989. Vols. 1 & 2

    NASA Technical Reports Server (NTRS)

    El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

    1992-01-01

    The present conference discusses such space nuclear power (SNP) issues as current design trends for SDI applications, ultrahigh heat-flux systems with curved surface subcooled nucleate boiling, design and manufacturing alternatives for low cost production of SNPs, a lightweight radioisotope heater for the Galileo mission, compatible materials for uranium fluoride-based gas core SNPs, Johnson noise thermometry for SNPs, and uranium nitride/rhenium compatibility studies for the SP-100 SNP. Also discussed are system issues in antimatter energy conversion, the thermal design of a heat source for a Brayton cycle radioisotope power system, structural and thermal analyses of an isotope heat source, a novel plant protection strategy for transient reactors, and beryllium toxicity.

  14. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  15. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  16. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

    1984-01-01

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  17. 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.

  18. A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Şahin, Sümer; Übeyli, Mustafa

    2008-12-01

    Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can't withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.

  19. Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

    Olson, D.L.

    1992-01-01

    The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

  20. Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

    Olson, D.L.

    1992-05-01

    The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory`s Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

  1. International-Aerial Measuring System (I-AMS) Training Program

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

    Wasiolek, Piotre T.; Malchor, Russell L.; Maurer, Richard J.

    2015-10-01

    Since the Fukushima reactor accident in 2011, there has been an increased interest worldwide in developing national capabilities to rapidly map and assess ground contamination resulting from nuclear reactor accidents. The capability to rapidly measure the size of the contaminated area, determine the activity level, and identify the radionuclides can aid emergency managers and decision makers in providing timely protective action recommendations to the public and first responders. The development of an aerial detection capability requires interagency coordination to assemble the radiation experts, detection system operators, and aviation aircrews to conduct the aerial measurements, analyze and interpret the data, andmore » provide technical assessments. The Office of International Emergency Management and Cooperation (IEMC) at the U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) sponsors an International - Aerial Measuring System (I-AMS) training program for partner nations to develop and enhance their response to radiological emergencies. An initial series of courses can be conducted in the host country to assist in developing an aerial detection capability. As the capability develops and expands, additional experience can be gained through advanced courses with the opportunity to conduct aerial missions over a broad range of radiation environments.« less

  2. Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994

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

    Lowendahl, B.

    1995-03-01

    In September 1992 Oskarshamn 1 was shut down in order to carry out measures to correct discovered deficiencies in the emergency cooling systems. Due to the results of a comprehensive non destructive test programme it was decided to perform a major replacement of pipes in the primary systems including a full system decontamination using the Siemens CORD process. The paper briefly presents the satisfying result of the decontamination performed in May-June 1993. When in late June 1993 cracks also were detected in the feed-water pipes situated inside the reactor pressure vessel (RPV) the plans were reconsidered and a large projectmore » was formed with the aim, in a first phase, to verify the integrity of the RPV. In order to make it possible to perform work manually inside the RPV special radiation protection measures had to be carried out. In January 1994 the lower region of the RPV was decontaminated, again using the CORD-process, followed by the installation of a special shielding construction in the RPV. The surprisingly good results of these efforts are also briefly described in the paper.« less

  3. Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

    NASA Astrophysics Data System (ADS)

    Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

    2017-02-01

    In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

  4. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  5. Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

    NASA Astrophysics Data System (ADS)

    Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

    2017-07-01

    In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

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

    NONE

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« less

  7. Selection criteria for wastewater treatment technologies to protect drinking water.

    PubMed

    von Sperling, M

    2000-01-01

    The protection of water bodies used as sources for drinking water is intimately linked to the adoption of adequate technologies for the treatment of the wastewater generated in the catchment area. The paper presents a general overview of the main technologies used for the treatment of domestic sewage, with a special emphasis on developing countries, and focussing on the main parameters of interest, such as BOD, coliforms and nutrients. A series of tables, figures and charts that can be used for the preliminary selection of treatment technologies is presented. The systems analysed are: stabilisation ponds, activated sludge, trickling filters, anaerobic systems and land disposal. Within each system, the main process variants are covered. Two summary tables are presented, one for quantitative analysis, including easily usable information based on per capita values (US$/cap, Watts/cap, m2 area/cap, m3 sludge/cap), and another for a qualitative comparison among the technologies, based on a one-to-five-star scoring system. The recent trend in tropical countries in the use of UASB (Upflow Anaerobic Sludge Blanket) reactors is also discussed.

  8. A liquid-metal filling system for pumped primary loop space reactors

    NASA Astrophysics Data System (ADS)

    Crandall, D. L.; Reed, W. C.

    Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

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

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

    Trond Bjornard; John Hockert

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

  10. Cost-Effective Systems for Atomic Layer Deposition

    ERIC Educational Resources Information Center

    Lubitz, Michael; Medina, Phillip A., IV; Antic, Aleks; Rosin, Joseph T.; Fahlman, Bradley D.

    2014-01-01

    Herein, we describe the design and testing of two different home-built atomic layer deposition (ALD) systems for the growth of thin films with sub-monolayer control over film thickness. The first reactor is a horizontally aligned hot-walled reactor with a vacuum purging system. The second reactor is a vertically aligned cold-walled reactor with a…

  11. 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...

  12. Assessment of the high temperature fission chamber technology for the French fast reactor program

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

    Jammes, C.; Filliatre, P.; Geslot, B.

    2011-07-01

    High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

  13. Trench fast reactor design using the microcomputer

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

    Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

    1987-01-01

    This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

  14. Analysis of boron dilution in a four-loop PWR

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

    Sun, J.G.; Sha, W.T.

    1995-03-01

    Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant ismore » at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.« less

  15. 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.

  16. MEANS FOR SHIELDING AND COOLING REACTORS

    DOEpatents

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

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  17. Exploratory study of several advanced nuclear-MHD power plant systems.

    NASA Technical Reports Server (NTRS)

    Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

    1973-01-01

    In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

  18. 77 FR 21594 - Duke Energy Carolinas, LLC; Environmental Assessment and Finding of No Significant Impact, Oconee...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-10

    ... of light-water nuclear power reactors provide adequate margins of safety during any condition of... toughness requirements for protection against pressurized thermal shock (PTS) events. The proposed action... for Protection Against Pressurized Thermal Shock Events,'' and 10 CFR part 50 Appendix G, ``Fracture...

  19. 10 CFR 140.22 - Commission guarantee and reimbursement agreements.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... required to have and maintain financial protection for each nuclear reactor as determined in § 140.11(a)(4... 10 Energy 2 2014-01-01 2014-01-01 false Commission guarantee and reimbursement agreements. 140.22 Section 140.22 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND...

  20. 10 CFR 140.22 - Commission guarantee and reimbursement agreements.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... required to have and maintain financial protection for each nuclear reactor as determined in § 140.11(a)(4... 10 Energy 2 2012-01-01 2012-01-01 false Commission guarantee and reimbursement agreements. 140.22 Section 140.22 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND...

  1. 10 CFR 140.22 - Commission guarantee and reimbursement agreements.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... required to have and maintain financial protection for each nuclear reactor as determined in § 140.11(a)(4... 10 Energy 2 2011-01-01 2011-01-01 false Commission guarantee and reimbursement agreements. 140.22 Section 140.22 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND...

  2. 10 CFR 140.22 - Commission guarantee and reimbursement agreements.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... required to have and maintain financial protection for each nuclear reactor as determined in § 140.11(a)(4... 10 Energy 2 2010-01-01 2010-01-01 false Commission guarantee and reimbursement agreements. 140.22 Section 140.22 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND...

  3. 10 CFR 140.22 - Commission guarantee and reimbursement agreements.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... required to have and maintain financial protection for each nuclear reactor as determined in § 140.11(a)(4... 10 Energy 2 2013-01-01 2013-01-01 false Commission guarantee and reimbursement agreements. 140.22 Section 140.22 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL PROTECTION REQUIREMENTS AND...

  4. Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.

    1988-01-01

    The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

  5. Analysis of boron dilution in a four-loop PWR

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

    Sun, J.G.; Sha, W.T.

    1995-12-31

    Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plantmore » is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.« less

  6. Remaining Sites Verification Package for the 118-C-3:3, 105-C French Drains, Waste Site Reclassification Form 2006-016

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

    R. A. Carlson

    2006-04-24

    The 118-C-3:3 french drains received condensate from the steam heating system in the 105-C Reactor Building. The 118-C-3:3 french drain meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

  7. Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

    NASA Astrophysics Data System (ADS)

    Abulfaraj, Waleed H.; Kamal, Salah M.

    1994-07-01

    The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concretes shielding. Multiattribute utility theory is selected to accommodate decision makers' preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Ilmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Ilmenite Serpentine concrete.

  8. The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

    Roglans, J; Hill, D J

    1994-02-01

    The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« less

  9. SNAP 10A FS-3 reactor performance

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

    Hawley, J.P.; Johnson, R.A.

    1966-08-15

    SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

  10. Weld monitor and failure detector for nuclear reactor system

    DOEpatents

    Sutton, Jr., Harry G.

    1987-01-01

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  11. Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

    NASA Astrophysics Data System (ADS)

    Darmawan, R.

    2018-01-01

    Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

  12. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

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

    Lescop, B.; Badeau, G.; Ivanovic, S.

    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. Themore » facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also modify the settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the broadcast of some particular information. At the guest institution, the information is displayed in two parts or screens, as shown in the Figure 3. Concerning the interaction with - and the feedback from - the remote classroom, the camera of the video system in the remote classroom is used to ensure the contact between the trainees and the lecturer and reactor operators. Thus, the Internet Reactor Laboratory is complementary to the on reactor training courses. It allows distant learning, reducing the overall cost of the course when this is necessary. It can efficiently be used for the development of the human resources needed by the nuclear industry and the nuclear programs in countries without research reactors.« less

  13. LFR "Lead-Cooled Fast Reactor"

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

    Cinotti, L; Fazio, C; Knebel, J

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small,more » transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology development already being carried out in different institutes participating in this STREP. This is particularly true in Europe where a large R&D program associated with the development of Accelerator Driven Systems (ADS) is being actively pursued. The general objective of the ELSY project is to design an innovative lead-cooled fast reactor complemented by an analytical effort to assess the existing knowledge base in the field of lead-alloy coolants (i.e., lead-bismuth eutectic (LBE) and also lead/lithium) in order to extrapolate this knowledge base to pure lead. This analysis effort will be complemented with some limited R&D activities to acquire missing or confirmatory information about fundamental topics for ELSY that are not sufficiently covered in the ongoing European ADS program or elsewhere.« less

  14. Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

    NASA Astrophysics Data System (ADS)

    Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

    2017-11-01

    In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

  15. Modeling and simulation of CANDU reactor and its regulating system

    NASA Astrophysics Data System (ADS)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

  16. 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.

  17. Reactor Operations Monitoring System

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

    Hart, M.M.

    1989-01-01

    The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

  18. Control of autothermal reforming reactor of diesel fuel

    NASA Astrophysics Data System (ADS)

    Dolanc, Gregor; Pregelj, Boštjan; Petrovčič, Janko; Pasel, Joachim; Kolb, Gunther

    2016-05-01

    In this paper a control system for autothermal reforming reactor for diesel fuel is presented. Autothermal reforming reactors and the pertaining purification reactors are used to convert diesel fuel into hydrogen-rich reformate gas, which is then converted into electricity by the fuel cell. The purpose of the presented control system is to control the hydrogen production rate and the temperature of the autothermal reforming reactor. The system is designed in such a way that the two control loops do not interact, which is required for stable operation of the fuel cell. The presented control system is a part of the complete control system of the diesel fuel cell auxiliary power unit (APU).

  19. A Review of Study on Thermal Energy Transport System by Synthesis and Decomposition Reactions of Methanol

    NASA Astrophysics Data System (ADS)

    Liu, Qiusheng; Yabe, Akira; Kajiyama, Shiro; Fukuda, Katsuya

    The study on thermal energy transport system by synthesis and decomposition reactions of methanol was reviewed. To promote energy conservation and global environment protection, a two-step liquid-phase methanol synthesis process, which starts with carbonylation of methanol to methyl formate, then followed by the hydrogenolysis of the formate, was studied to recover wasted or unused discharged heat from industrial sources for the thermal energy demands of residential and commercial areas by chemical reactions. The research and development of the system were focused on the following three points. (1) Development of low-temperature decomposition and synthetic catalysts, (2) Development of liquid phase reactor (heat exchanger accompanying chemical reaction), (3) Simulation of the energy transport efficiency of entire system which contains heat recovery and supply sections. As the result of the development of catalyst, promising catalysts which agree with the development purposes for the methyl formate decomposition reaction and the synthetic reaction are being developed though some studies remain for the methanol decomposition and synthetic reactions. In the fundamental development of liquid phase reactor, the solubilities of CO and H2 gases in methanol and methyl formate were measured by the method of total pressure decrease due to absorption under pressures up to 1500kPa and temperatures up to 140°C. The diffusivity of CO gas in methanol was determined by measuring the diameter and solution time of single CO bubbles in methanol. The chemical reaction rate of methanol synthesis by hydrogenolysis of methyl formate was measured using a plate-type of Raney copper catalyst in a reactor with rectangular channel and in an autoclave reactor. The reaction characteristics were investigated by carrying out the experiments at various temperatures, flow rates and at various catalyst development conditions. We focused on the effect of Raney copper catalyst thickness on the liquid-phase chemical reaction by varying the development time of the catalyst. Investigation results of the catalyst such as surface area, pore radius, lattice size, and photographs of scanning electron microscope (SEM) were also given. In the simulation of energy transport efficiency of this system, by simulating the energy transfer system using two-step liquid phase methanol decomposition and synthetic reactions, and comparing with the technology so far, it can be expected that an innovative energy transfer system is possible to realize.

  20. Demonstration of Robustness and Integrated Operation of a Series-Bosch System

    NASA Technical Reports Server (NTRS)

    Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

    2016-01-01

    Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

  1. Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation

    NASA Astrophysics Data System (ADS)

    Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François

    2017-12-01

    A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

  2. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

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

    1959-11-01

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

  3. R and D program for core instrumentation improvements devoted for French sodium fast reactors

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

    Jeannot, J. P.; Rodriguez, G.; Jammes, C.

    2011-07-01

    Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within themore » scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)« less

  4. Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

    PubMed

    Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

    2018-01-04

    A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

  5. Oxygenic photosynthesis as a protection mechanism for cyanobacteria against iron-encrustation in environments with high Fe2+ concentrations

    PubMed Central

    Ionescu, Danny; Buchmann, Bettina; Heim, Christine; Häusler, Stefan; de Beer, Dirk; Polerecky, Lubos

    2014-01-01

    If O2 is available at circumneutral pH, Fe2+ is rapidly oxidized to Fe3+, which precipitates as FeO(OH). Neutrophilic iron oxidizing bacteria have evolved mechanisms to prevent self-encrustation in iron. Hitherto, no mechanism has been proposed for cyanobacteria from Fe2+-rich environments; these produce O2 but are seldom found encrusted in iron. We used two sets of illuminated reactors connected to two groundwater aquifers with different Fe2+ concentrations (0.9 μM vs. 26 μM) in the Äspö Hard Rock Laboratory (HRL), Sweden. Cyanobacterial biofilms developed in all reactors and were phylogenetically different between the reactors. Unexpectedly, cyanobacteria growing in the Fe2+-poor reactors were encrusted in iron, whereas those in the Fe2+-rich reactors were not. In-situ microsensor measurements showed that O2 concentrations and pH near the surface of the cyanobacterial biofilms from the Fe2+-rich reactors were much higher than in the overlying water. This was not the case for the biofilms growing at low Fe2+ concentrations. Measurements with enrichment cultures showed that cyanobacteria from the Fe2+-rich environment increased their photosynthesis with increasing Fe2+ concentrations, whereas those from the low Fe2+ environment were inhibited at Fe2+ > 5 μM. Modeling based on in-situ O2 and pH profiles showed that cyanobacteria from the Fe2+-rich reactor were not exposed to significant Fe2+ concentrations. We propose that, due to limited mass transfer, high photosynthetic activity in Fe2+-rich environments forms a protective zone where Fe2+ precipitates abiotically at a non-lethal distance from the cyanobacteria. This mechanism sheds new light on the possible role of cyanobacteria in precipitation of banded iron formations. PMID:25228899

  6. Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment

    DOEpatents

    Grossman, Mark W.

    1991-01-01

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

  7. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

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

    1974-01-01

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

  8. Enhanced protection of PDMS-embedded palladium catalysts by co-embedding of sulphide-scavengers.

    PubMed

    Comandella, Daniele; Ahn, Min Hyung; Kim, Hojeong; Mackenzie, Katrin

    2017-12-01

    For Pd-containing hydrodechlorination catalysts, coating with poly(dimethyl siloxane) (PDMS) was proposed earlier as promising protection scheme against poisoning. The PDMS coating can effectively repel non-permeating poisons (such as SO 3 2- ) retaining the hydrodechlorination Pd activity. In the present study, the previously achieved protection efficiency was enhanced by incorporation of sulphide scavengers into the polymer. The embedded scavengers were able to bind permeating non-ionic poisons (such as H 2 S) during their passage through PDMS prior to Pd contact which ensured an extended catalyst lifetime. Three scavenger types forming non-permeable sulphur species from H 2 S - alkaline, oxidative or iron-based compounds - were either incorporated into single-layer coats around individual Pd/Al 2 O 3 particles or into a second layer above Pd-containing PDMS films (Pd-PDMS). Hydrodechlorination and hydrogenation were chosen as model reactions, carried out in batch and continuous-flow reactors. Batch tests with all scavenger-containing catalysts showed extended Pd protection compared to scavenger-free catalysts. Solid alkaline compounds (Ca(OH) 2 , NaOH, CaO) and MnO 2 showed the highest instantaneous scavenger efficiencies (retained Pd activity=30-60%), while iron-based catalysts, such as nano zero-valent iron (nZVI) or ferrocene (FeCp 2 ), proved less efficient (1-10%). When stepwise poisoning was applied, the protection efficiency of iron-based and oxidizing compounds was higher in the long term than that of alkaline solids. Long-term experiments in mixed-flow reactors were performed with selected scavengers, revealing the following trend of protection efficiency: CaO 2 >Ca(OH) 2 >FeCp 2 . Under field-simulating conditions using a fixed-bed reactor, the combination of sulphide pre-oxidation in the water phase by H 2 O 2 and local scavenger-enhanced Pd protection was successful. The oxidizing agent H 2 O 2 does not disturb the Pd-catalysed reduction, while the PDMS-incorporated scavenger considerably extends the catalyst life in the presence of H 2 S. This work demonstrates that the scavenger-based protection strategy is an effective means to increase the resistance of PDMS-embedded Pd against permeating poisons. Copyright © 2017. Published by Elsevier B.V.

  9. 76 FR 80410 - Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Radiation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-23

    ... Compliance with Packaging Requirements for Shipment and Receipt of Radioactive Material.'' The Subcommittee... Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on January 18, 2012, Room T-2B3, 11545...

  10. 10 CFR 72.10 - Employee protection.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Employee protection. 72.10 Section 72.10 Energy NUCLEAR..., HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.10... Form 3, “Notice to Employees,” referenced in 10 CFR 19.11(c). This form must be posted at locations...

  11. 10 CFR 72.10 - Employee protection.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Employee protection. 72.10 Section 72.10 Energy NUCLEAR..., HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.10... Form 3, “Notice to Employees,” referenced in 10 CFR 19.11(c). This form must be posted at locations...

  12. Portable thermo-photovoltaic power source

    DOEpatents

    Zuppero, Anthony C.; Krawetz, Barton; Barklund, C. Rodger; Seifert, Gary D.

    1997-01-14

    A miniature thermo-photovoltaic (TPV) device for generation of electrical power for use in portable electronic devices. A TPV power source is constructed to provide a heat source chemical reactor capable of using various fuels, such as liquid hydrocarbons, including but not limited to propane, LPG, butane, alcohols, oils and diesel fuels to generate a source of photons. A reflector dish guides misdirected photon energy from the photon source toward a photovoltaic array. A thin transparent protector sheet is disposed between the photon source and the array to reflect back thermal energy that cannot be converted to electricity, and protect the array from thermal damage. A microlens disposed between the protector sheet and the array further focuses the tailored band of photon energy from the photon source onto an array of photovoltaic cells, whereby the photon energy is converted to electrical power. A heat recuperator removes thermal energy from reactor chamber exhaust gases, preferably using mini- or micro-bellows to force air and fuel past the exhaust gases, and uses the energy to preheat the fuel and oxidant before it reaches the reactor, increasing system efficiency. Mini- or micro-bellows force ambient air through the system both to supply oxidant and to provide cooling. Finally, an insulator, which is preferably a super insulator, is disposed around the TPV power source to reduce fuel consumption, and to keep the TPV power source cool to the touch so it can be used in hand-held devices.

  13. Tools for placing the radiological health hazard in perspective following a severe emergency at a light water reactor (LWR) or its spent fuel pool.

    PubMed

    McKenna, Thomas; Welter, Phillip Vilar; Callen, Jessica; Martincic, Rafael; Dodd, Brian; Kutkov, Vladimir

    2015-01-01

    Experience from past nuclear and radiological emergencies shows that placing the radiological health hazard in perspective and having a definition of "safe" are required in order to prevent members of the public, those responsible for protecting the public (i.e., decision makers), and others from taking inappropriate and damaging actions that are not justified based on the radiological health hazard. The principle concerns of the public during a severe nuclear power plant or spent fuel pool emergency are "Am I safe?" and "What should I do to be safe?" However, these questions have not been answered to the satisfaction of the public, despite various protective actions being implemented to ensure their safety. Instead, calculated doses or various measured quantities (e.g., ambient dose rate or radionuclide concentrations) are used to describe the situation to the public without placing them into perspective in terms of the possible radiological health hazard, or if they have, it has been done incorrectly. This has contributed to members of the public taking actions that do more harm than good in the belief that they are protecting themselves. Based on established international guidance, this paper provides a definition of "safe" for the radiological health hazard for use in nuclear or radiological emergencies and a system for putting the radiological health hazard in perspective for quantities most commonly measured after a release resulting from a severe emergency at a light water reactor or its spent fuel pool.

  14. U.S. Department of Energy physical protection upgrades at the Latvian Academy of Sciences Nuclear Research Center, Latvia

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

    Haase, M.; Hine, C.; Robertson, C.

    1996-12-31

    Approximately five years ago, the Safe, Secure Dismantlement program was started between the US and countries of the Former Soviet Union (FSU). The purpose of the program is to accelerate progress toward reducing the risk of nuclear weapons proliferation, including such threats as theft, diversion, and unauthorized possession of nuclear materials. This would be accomplished by strengthening the material protection, control, and accounting systems within the FSU countries. Under the US Department of Energy`s program of providing cooperative assistance to the FSU countries in the areas of Material Protection, Control, and Accounting (MPC and A), the Latvian Academy of Sciencesmore » Nuclear Research Center (LNRC) near Riga, Latvia, was identified as a candidate site for a cooperative MPC and A project. The LNRC is the site of a 5-megawatt IRT-C pool-type research reactor. This paper describes: the process involved, from initial contracting to project completion, for the physical protection upgrades now in place at the LNRC; the intervening activities; and a brief overview of the technical aspects of the upgrades.« less

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

    Adams, S.R.

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

  16. Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.

    1988-01-01

    A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

  17. 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.

  18. 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.

  19. Integrated shielding systems for manned interplanetary spaceflight

    NASA Astrophysics Data System (ADS)

    George, Jeffrey A.

    1992-01-01

    The radiation environment encountered by manned interplanetary missions can have a severe impact on both vehicle design and mission performance. This study investigates the potential impact of radiation protection on interplanetary vehicle design for a manned Mars mission. A systems approach was used to investigate the radiation protection requirements of the sum interplanetary environment. Radiation budgets were developed which result in minimum integrated shielding system masses for both nuclear and non-nuclear powered missions. A variety of system configurations and geometries were assessed over a range of dose constraints. For an annual dose equivalent rate limit of 50 rem/yr, an environmental shielding system composed of a habitat shield and storm shelter was found to result in the lowest total mass. For a limit of 65 rem/yr, a system composed of a sleeping quarters shield was least massive, and resulted in significantly reduced system mass. At a limit of 75 rem/yr, a storm shelter alone was found to be sufficient, and exhibited a further mass reduction. Optimal shielding system results for 10 MWe nuclear powered missions were found to follow along similar lines, with the addition of a reactor shadow shield. A solar minimum galactic cosmic ray spectrum and one anomalously large solar particle event during the course of a two year mission were assumed. Water was assumed for environmental radiation shielding.

  20. Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

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

    Wright, Steven A.; Sanchez, Travis

    2005-02-06

    The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

  1. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  2. Summary of NR Program Prometheus Efforts

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

    J Ashcroft; C Eshelman

    2006-02-08

    The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

  3. High Solid Fed-batch Butanol Fermentation with Simultaneous Product Recovery: Part II - Process Integration.

    PubMed

    Qureshi, Nasib; Klasson, K Thomas; Saha, Badal C; Liu, Siqing

    2018-04-25

    In these studies liquid hot water (LHW) pretreated and enzymatically hydrolyzed Sweet Sorghum Bagasse (SSB) hydrolyzates were fermented in a fed-batch reactor. As reported in the preceding paper, the culture was not able to ferment the hydrolyzate I in a batch process due to presence of high level of toxic chemicals, in particular acetic acid released from SSB during the hydrolytic process. To be able to ferment the hydrolyzate I obtained from 250 gL -1 SSB hydrolysis, a fed-batch reactor with in-situ butanol recovery was devised. The process was started with the hydrolyzate II and when good cell growth and vigorous fermentation were observed, the hydrolyzate I was slowly fed to the reactor. In this manner the culture was able to ferment all the sugars present in both the hydrolyzates to acetone butanol ethanol (ABE). In a control batch reactor in which ABE was produced from glucose, ABE productivity and yield of 0.42 gL -1 h -1 and 0.36 were obtained, respectively. In the fed-batch reactor fed with SSB hydrolyzates these productivity and yield values were 0.44 gL -1 h -1 and 0.45, respectively. ABE yield in the integrated system was high due to utilization of acetic acid to convert to ABE. In summary we were able to utilize both the hydrolyzates obtained from LHW pretreated and enzymatically hydrolyzed SSB (250 gL -1 ) and convert them to ABE. Complete fermentation was possible due to simultaneous recovery of ABE by vacuum. This article is protected by copyright. All rights reserved. © 2018 American Institute of Chemical Engineers.

  4. SP-100 Control System Design

    NASA Astrophysics Data System (ADS)

    Shukla, Jaikaran N.; Halfen, Frank J.; Brynsvold, Glen V.; Syed, Akbar; Jiang, Thomas J.; Wong, Kwok K.; Otwell, Robert L.

    1994-07-01

    Recent work in lower power generic early applications for the SP-100 have resulted in control system design simplification for a 20 kWe design with thermoelectric power conversion. This paper presents the non-mission-dependent control system features for this design. The control system includes a digital computer based controller, dual purpose control rods and drives, temperature sensors, and neutron flux monitors. The thaw system is mission dependent and can be either electrical or based on NaK trace lines. Key features of the control system and components are discussed. As was the case for higher power applications, the initial on-orbit approach to criticality involves the relatively fast withdrawal of the control-rods to a near-critical position followed by slower movement through critical and into the power range. The control system performs operating maneuvers as well as providing for automatic startup, shutdown, restart, and reactor protection.

  5. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

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

  6. Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

    NASA Technical Reports Server (NTRS)

    Clark, John S. (Editor)

    1991-01-01

    Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

  7. PBF Reactor Building (PER620). Camera faces south along west wall. ...

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

    PBF Reactor Building (PER-620). Camera faces south along west wall. Gap between native lava rock and concrete basement walls is being backfilled and compacted. Wire mesh protects workers from falling rock. Note penetrations for piping that will carry secondary coolant water to Cooling Tower. Photographer: Holmes. Date: June 15, 1967. INEEL negative no. 67-3665 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  8. Safety approach to the selection of design criteria for the CRBRP reactor refueling system

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

    Meisl, C J; Berg, G E; Sharkey, N F

    1979-01-01

    The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

  9. Axi-symmetrical flow reactor for [sup 196]Hg photochemical enrichment

    DOEpatents

    Grossman, M.W.

    1991-04-30

    The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, [sup 196]Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired [sup 196]Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith. 10 figures.

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

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

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

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

    Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

    The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

  12. 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.

  13. Simultaneous saccharification and extractive fermentation of lignocellulosic materials into lactic acid in a two-zone fermentor-extractor system.

    PubMed

    Iyer, P V; Lee, Y Y

    1999-01-01

    Simultaneous saccharification and extractive fermentation of lignocellulosic materials into lactic acid was investigated using a two-zone bioreactor. The system is composed of an immobilized cell reactor, a separate column reactor containing the lignocellulosic substrate and a hollow-fiber membrane. It is operated by recirculating the cell free enzyme (cellulase) solution from the immobilized cell reactor to the column reactor through the membrane. The enzyme and microbial reactions thus occur at separate locations, yet simultaneously. This design provides flexibility in reactor operation as it allows easy separation of the solid substrate from the microorganism, in situ removal of the product and, if desired, different temperatures in the two reactor sections. This reactor system was tested using pretreated switchgrass as the substrate. It was operated under a fed-batch mode with continuous removal of lactic acid by solvent extraction. The overall lactic acid yield obtainable from this bioreactor system is 77% of the theoretical.

  14. Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

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

    Chang, L.K.; Mohr, D.; Planchon, H.P.

    This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

  15. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    NASA Astrophysics Data System (ADS)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  16. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

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

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

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less

  17. Results of detailed analyses performed on boring cores extracted from the concrete floors of the Fukushima Daiichi nuclear power plant reactor buildings

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

    Maeda, Koji; Sasaki, S.; Kumai, M.

    Due to the massive earthquake and tsunami on March 11, 2011, and the following severe accident at the Fukushima Daiichi Nuclear Power Plant, concrete surfaces within the reactor buildings were exposed to radioactive liquid and vapor phase contaminants. In order to clarify the situation of this contamination in the reactor buildings of Units 1, 2 and 3, selected samples were transported to the Fuels Monitoring Facility in the Oarai Engineering Center of JAEA where they were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. In particular, penetration of radiocesium inmore » the surface coatings layer and sub-surface concrete was evaluated. The analysis results indicate that the situation of contamination in the building of Unit 2 was different from others, and the protective surface coatings on the concrete floors provided significant protection against radionuclide penetration. The localized penetration of contamination in the concrete floors was found to be confined within a millimeter of the surface of the coating layer of some millimeters. (authors)« less

  18. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  19. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

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

  20. Seismic attenuation system for a nuclear reactor

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

    Liszkai, Tamas; Cadell, Seth

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vesselmore » are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.« less

  1. Cooling system for a nuclear reactor

    DOEpatents

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  2. A coupled nuclear reactor thermal energy storage system for enhanced load following operation

    NASA Astrophysics Data System (ADS)

    Alameri, Saeed A.

    Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

  3. Nuclear reactor cavity floor passive heat removal system

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

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

  4. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  5. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  6. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

    DOEpatents

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

    2015-12-29

    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  7. AP1000{sup R} severe accident features and post-Fukushima considerations

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

    Scobel, J. H.; Schulz, T. L.; Williams, M. G.

    2012-07-01

    The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, themore » AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)« less

  8. Radiation Transport Calculation of the UGXR Collimators for the Jules Horowitz Reactor (JHR)

    NASA Astrophysics Data System (ADS)

    Chento, Yelko; Hueso, César; Zamora, Imanol; Fabbri, Marco; Fuente, Cristina De La; Larringan, Asier

    2017-09-01

    Jules Horowitz Reactor (JHR), a major infrastructure of European interest in the fission domain, will be built and operated in the framework of an international cooperation, including the development and qualification of materials and nuclear fuel used in nuclear industry. For this purpose UGXR Collimators, two multi slit gamma and X-ray collimation mechatronic systems, will be installed at the JHR pool and at the Irradiated Components Storage pool. Expected amounts of radiation produced by the spent fuel and X-ray accelerator implies diverse aspects need to be verified to ensure adequate radiological zoning and personnel radiation protection. A computational methodology was devised to validate the Collimators design by means of coupling different engineering codes. In summary, several assessments were performed by means of MCNP5v1.60 to fulfil all the radiological requirements in Nominal scenario (TEDE < 25µSv/h) and in Maintenance scenario (TEDE < 2mSv/h) among others, detailing the methodology, hypotheses and assumptions employed.

  9. 78 FR 34677 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-10

    ... ensure adequate protection of public health and safety, to promote the common defense and security, and to protect the environment. One way to support this mission is through the implementation of the Reactor Oversight Process (ROP), which is the agency's program to inspect, measure, and assess the safety...

  10. The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

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

    Fanning, T. H.; Brunett, A. J.; Sumner, T.

    The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

  11. Process for photosynthetically splitting water

    DOEpatents

    Greenbaum, Elias

    1984-01-01

    The invention is an improved process for producing gaseous hydrogen and oxygen from water. The process is conducted in a photolytic reactor which contains a water-suspension of a photoactive material containing a hydrogen-liberating catalyst. The reactor also includes a volume for receiving gaseous hydrogen and oxygen evolved from the liquid phase. To avoid oxygen-inactivation of the catalyst, the reactor is evacuated continuously by an external pump which circulates the evolved gases through means for selectively recovering hydrogen therefrom. The pump also cools the reactor by evaporating water from the liquid phase. Preferably, product recovery is effected by selectively diffusing the hydrogen through a heated semipermeable membrane, while maintaining across the membrane a magnetic field gradient which biases the oxygen away from the heated membrane. This promotes separation, minimizes the back-reaction of hydrogen and oxygen, and protects the membrane.

  12. NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL ASSOCIATED WITH A CLOSED FUEL CYCLE

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

    Bathke, C. G.; Ebbinghaus, B.; Sleaford, Brad W.

    2010-06-11

    This paper examines the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the various processing steps required for a closed fuel cycle. This paper combines the results from earlier studies that examined the attractiveness of SNM associated with the processing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR with new results for the final, repeated burning of SNM in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). The results of this paper suggest that all reprocessing products evaluated so farmore » need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, or undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.« less

  13. LLNL Results from CALIBAN-PROSPERO Nuclear Accident Dosimetry Experiments in September 2014

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

    Lobaugh, M. L.; Hickman, D. P.; Wong, C. W.

    2015-05-21

    Lawrence Livermore National Laboratory (LLNL) uses thin neutron activation foils, sulfur, and threshold energy shielding to determine neutron component doses and the total dose from neutrons in the event of a nuclear criticality accident. The dosimeter also uses a DOELAP accredited Panasonic UD-810 (Panasonic Industrial Devices Sales Company of America, 2 Riverfront Plaza, Newark, NJ 07102, U.S.A.) thermoluminescent dosimetery system (TLD) for determining the gamma component of the total dose. LLNL has participated in three international intercomparisons of nuclear accident dosimeters. In October 2009, LLNL participated in an exercise at the French Commissariat à l’énergie atomique et aux énergies alternativesmore » (Alternative Energies and Atomic Energy Commission- CEA) Research Center at Valduc utilizing the SILENE reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison at CEA Valduc, this time with exposures at the CALIBAN reactor (Hickman et al. 2011). This paper discusses LLNL’s results of a third intercomparison hosted by the French Institut de Radioprotection et de Sûreté Nucléaire (Institute for Radiation Protection and Nuclear Safety- IRSN) with exposures at two CEA Valduc reactors (CALIBAN and PROSPERO) in September 2014. Comparison results between the three participating facilities is presented elsewhere (Chevallier 2015; Duluc 2015).« less

  14. Reactor application of an improved bundle divertor

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

    Yang, T.F.; Ruck, G.W.; Lee, A.Y.

    1978-11-01

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

  15. Vessel thermal map real-time system for the JET tokamak

    NASA Astrophysics Data System (ADS)

    Alves, D.; Felton, R.; Jachmich, S.; Lomas, P.; McCullen, P.; Neto, A.; Valcárcel, D. F.; Arnoux, G.; Card, P.; Devaux, S.; Goodyear, A.; Kinna, D.; Stephen, A.; Zastrow, K.-D.

    2012-05-01

    The installation of international thermonuclear experimental reactor-relevant materials for the plasma facing components (PFCs) in the Joint European Torus (JET) is expected to have a strong impact on the operation and protection of the experiment. In particular, the use of all-beryllium tiles, which deteriorate at a substantially lower temperature than the formerly installed carbon fiber composite tiles, imposes strict thermal restrictions on the PFCs during operation. Prompt and precise responses are therefore required whenever anomalous temperatures are detected. The new vessel thermal map real-time application collects the temperature measurements provided by dedicated pyrometers and infrared cameras, groups them according to spatial location and probable offending heat source, and raises alarms that will trigger appropriate protective responses. In the context of the JET global scheme for the protection of the new wall, the system is required to run on a 10 ms cycle communicating with other systems through the real-time data network. In order to meet these requirements a commercial off-the-shelf solution has been adopted based on standard x86 multicore technology. Linux and the multithreaded application real-time executor (MARTe) software framework were respectively the operating system of choice and the real-time framework used to build the application. This paper presents an overview of the system with particular technical focus on the configuration of its real-time capability and the benefits of the modular development approach and advanced tools provided by the MARTe framework.

  16. Particle Filter-Based Recursive Data Fusion With Sensor Indexing for Large Core Neutron Flux Estimation

    NASA Astrophysics Data System (ADS)

    Tamboli, Prakash Kumar; Duttagupta, Siddhartha P.; Roy, Kallol

    2017-06-01

    We introduce a sequential importance sampling particle filter (PF)-based multisensor multivariate nonlinear estimator for estimating the in-core neutron flux distribution for pressurized heavy water reactor core. Many critical applications such as reactor protection and control rely upon neutron flux information, and thus their reliability is of utmost importance. The point kinetic model based on neutron transport conveniently explains the dynamics of nuclear reactor. The neutron flux in the large core loosely coupled reactor is sensed by multiple sensors measuring point fluxes located at various locations inside the reactor core. The flux values are coupled to each other through diffusion equation. The coupling facilitates redundancy in the information. It is shown that multiple independent data about the localized flux can be fused together to enhance the estimation accuracy to a great extent. We also propose the sensor anomaly handling feature in multisensor PF to maintain the estimation process even when the sensor is faulty or generates data anomaly.

  17. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  18. Kinetics of D-lactic acid production by Sporolactobacillus sp. strain CASD using repeated batch fermentation.

    PubMed

    Zhao, Bo; Wang, Limin; Li, Fengsong; Hua, Dongliang; Ma, Cuiqing; Ma, Yanhe; Xu, Ping

    2010-08-01

    D-lactic acid was produced by Sporolactobacillus sp. strain CASD in repeated batch fermentation with one- and two-reactor systems. The strain showed relatively high energy consumption in its growth-related metabolism in comparison with other lactic acid producers. When the fermentation was repeated with 10% (v/v) of previous culture to start a new batch, D-lactic acid production shifted from being cell-maintenance-dependent to cell-growth-dependent. In comparison with the one-reactor system, D-lactic acid production increased approximately 9% in the fourth batch of the two-reactor system. Strain CASD is an efficient D-lactic acid producer with increased growth rate at the early stage of repeated cycles, which explains the strain's physiological adaptation to repeated batch culture and improved performance in the two-reactor fermentation system. From a kinetic point of view, two-reactor fermentation system was shown to be an alternative for conventional one-reactor repeated batch operation. Copyright 2010 Elsevier Ltd. All rights reserved.

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

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

    Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

    2009-09-15

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

  20. Hydrogasification reactor and method of operating same

    DOEpatents

    Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

    2013-09-10

    The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

  1. 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.

  2. Improved vortex reactor system

    DOEpatents

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

    1995-05-09

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

  3. Field experience with remote monitoring

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

    Desrosiers, A.E.

    1995-03-01

    The Remote Monitoring System (RMS) is a combination of Merlin Gerin detection hardware, digital data communications hardware, and computer software from Bartlett Services, Inc. (BSI) that can improve the conduct of reactor plant operations in several areas. Using the RMS can reduce radiation exposures to radiation protection technicians (RPTs), reduce radiation exposures to plant maintenance and operations personnel, and reduce the time required to complete maintenance and inspections during outages. The number of temporary RPTs required during refueling outages can also be reduced. Data from use of the RMS at a two power plants are presented to illustrate these points.

  4. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (Editor)

    1973-01-01

    Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

  5. Exhaust system with emissions storage device and plasma reactor

    DOEpatents

    Hoard, John W.

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  6. 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.

  7. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  8. 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.

  9. WATER PUMP HOUSE, TRA619, AND TWO WATER STORAGE RESERVOIRS. INDUSTRIAL ...

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

    WATER PUMP HOUSE, TRA-619, AND TWO WATER STORAGE RESERVOIRS. INDUSTRIAL WINDOWS AND COPING STRIPS AT TOP OF WALLS AND ENTRY VESTIBULE. BOLLARDS PROTECT UNDERGROUND FACILITIES. SWITCHYARD AT RIGHT EDGE OF VIEW. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON ORIGINAL NEGATIVE. INL NEGATIVE NO. 3816. Unknown Photographer, 11/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  10. Novel duplex vapor: Electrochemical method for silicon solar cells. [chemical reactor for a silicon sodium reaction system

    NASA Technical Reports Server (NTRS)

    Nanis, L.; Sanjurjo, A.; Sancier, K.

    1979-01-01

    The scaled up chemical reactor for a SiF4-Na reaction system is examined for increased reaction rate and production rate. The reaction system which now produces 5 kg batches of mixed Si and NaF is evaluated. The reactor design is described along with an analysis of the increased capacity of the Na chip feeder. The reactor procedure is discussed and Si coalescence in the reaction products is diagnosed.

  11. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  12. 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...

  13. Application of a Self-Actuating Shutdown System (SASS) to a Gas-Cooled Fast Reactor (GCFR)

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

    Germer, J.H.; Peterson, L.F.; Kluck, A.L.

    1980-09-01

    The application of a SASS (Self-Actuated Shutdown System) to a GCFR (Gas-Cooled Fast Reactor) is compared with similar systems designed for an LMFBR (Liquid Metal Fast Breeder Reactor). A comparison of three basic SASS concepts is given: hydrostatic holdup, fluidic control, and magnetic holdup.

  14. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  15. A low-cost municipal sewage treatment system with a combination of UASB and the "fourth-generation" downflow hanging sponge reactors.

    PubMed

    Tandukar, M; Uemura, S; Machdar, I; Ohashi, A; Harada, H

    2005-01-01

    This paper presents an evaluation of the process performance of a pilot-scale "fourth generation" downflow hanging sponge (DHS) post-treatment system combined with a UASB pretreatment unit treating municipal wastewater. After the successful operation of the second- and third-generation DHS reactors, the fourth-generation DHS reactor was developed to overcome a few shortcomings of its predecessors. This reactor was designed to further enhance the treatment efficiency and simplify the construction process in real scale, especially for the application in developing countries. Configuration of the reactor was modified to enhance the dissolution of air into the wastewater and to avert the possible clogging of the reactor especially during sudden washout from the UASB reactor. The whole system was operated at a total hydraulic retention time (HRT) of 8 h (UASB: 6 h and DHS: 2 h) for a period of over 600 days. The combined system was able to remove 96% of unfiltered BOD with only 9 mg/L remaining in the final effluent. Likewise, F. coli were removed by 3.45 log with the final count of 10(3) to 10(4) MPN/100 ml. Nutrient removal by the system was also satisfactory.

  16. Multi-Megawatt Power System Trade Study

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

    Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

    2001-11-01

    As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

  17. 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.

  18. Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

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

    Ahmed, K. K.; Scarlat, R. O.; Hu, R.

    Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

  19. 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

  20. Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

    NASA Astrophysics Data System (ADS)

    Bragg-Sitton, Shannon M.

    The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

  1. An adaptive load-following control system for a space nuclear power system

    NASA Astrophysics Data System (ADS)

    Metzger, John D.; El-Genk, Mohamed S.

    An adaptive load-following control system is proposed for a space nuclear power system. The conceptual design of the SP-100 space nuclear power system proposes operating the nuclear reactor at a base thermal power and accommodating changes in the electrical power demand with a shunt regulator. It is necessary to increase the reactor thermal power if the payload electrical demand exceeds the peak system electrical output for the associated reactor power. When it is necessary to change the nuclear reactor power to meet a change in the power demand, the power ascension or descension must be accomplished in a predetermined manner to avoid thermal stresses in the system and to achieve the desired reactor period. The load-following control system described has the ability to adapt to changes in the system and to changes in the satellite environment. The application is proposed of the model reference adaptive control (MRAC). The adaptive control system has the ability to control the dynamic response of nonlinear systems. Three basic subsets of adaptive control are: (1) gain scheduling, (2) self-tuning regulators, and (3) model reference adaptive control.

  2. 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.

  3. A Comparison of Fission Power System Options for Lunar and Mars Surface Applications

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.

    2006-01-01

    This paper presents a comparison of reactor and power conversion design options for 50 kWe class lunar and Mars surface power applications with scaling from 25 to 200 kWe. Design concepts and integration approaches are provided for three reactor-converter combinations: gas-cooled Brayton, liquid-metal Stirling, and liquid-metal thermoelectric. The study examines the mass and performance of low temperature, stainless steel based reactors and higher temperature refractory reactors. The preferred system implementation approach uses crew-assisted assembly and in-situ radiation shielding via installation of the reactor in an excavated hole. As an alternative, self-deployable system concepts that use earth-delivered, on-board radiation shielding are evaluated. The analyses indicate that among the 50 kWe stainless steel reactor options, the liquid-metal Stirling system provides the lowest mass at about 5300 kg followed by the gas-cooled Brayton at 5700 kg and the liquid-metal thermoelectric at 8400 kg. The use of a higher temperature, refractory reactor favors the gas-cooled Brayton option with a system mass of about 4200 kg as compared to the Stirling and thermoelectric options at 4700 and 5600 kg, respectively. The self-deployed concepts with on-board shielding result in a factor of two system mass increase as compared to the in-situ shielded concepts.

  4. Fluidized bed coal combustion reactor

    NASA Technical Reports Server (NTRS)

    Moynihan, P. I.; Young, D. L. (Inventor)

    1981-01-01

    A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

  5. SP-100 Program: space reactor system and subsystem investigations

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

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  6. SP-100 program: Space reactor system and subsystem investigations

    NASA Astrophysics Data System (ADS)

    Harty, R. B.

    1983-09-01

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

  7. Survey of Materials for Fusion Fission Hybrid Reactors Vol 1 Rev. 0

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

    Farmer, Joseph Collin

    2007-07-03

    Materials for fusion-fission hybrid reactors fall into several broad categories, including fuels, blanket and coolant materials, cladding, structural materials, shielding, and in the specific case of inertial-confinement fusion systems, laser and optical materials. This report surveys materials in all categories of materials except for those required for lasers and optics. Preferred collants include two molten salt mixtures known as FLIBE (Li2BeF4) and FLINABE (LiNaBeF4). In the case of homogenous liquid fuels, UF4 can be dissolved in these molten salt mixtures. The transmutation of lithium in this coolant produces very corrosive hydrofluoric acid species (HF and TF), which can rapidly degrademore » structural materials. Broad ranges of high-melting radiation-tolerant structural material have been proposed for fusion-fission reactor structures. These include a wide variety of steels and refractory alloys. Ferritic steels with oxide-dispersion strengthening and graphite have been given particular attention. Refractory metals are found in Groups IVB and VB of the periodic table, and include Nb, Ta, Cr, Mo, and W, as serve as the basis of refractory alloys. Stable high-melting composites and amorphous metals may also be useful. Since amorphous metals have no lattice structure, neutron bombardment cannot dislodge atoms from lattice sites, and the materials would be immune from this specific mode of degradation. The free energy of formation of fluorides of the alloying elements found in steels and refractory alloys can be used to determine the relative stability of these materials in molten salts. The reduction of lithium transmutation products (H + and T +) drives the electrochemical corrosion process, and liberates aggressive fluoride ions that pair with ions formed from dissolved structural materials. Corrosion can be suppressed through the use of metallic Be and Li, though the molten salt becomes laden with colloidal suspensions of Be and Li corrosion products in the process. Alternatively, imposed currents and other high-temperature cathodic protection systems are envisioned for protection of the structural materials. This novel concept could prove to be enabling technology for such high-temperature molten-salt reactors. The use of UF 4 as a liquid-phase homogenous fuel is also complicated by redox control. For example, the oxidation of tetravalent uranium to hexavalent uranium could result in the formation of volatile UF 6. This too could be controlled through electrochemically manipulated oxidation and reduction reactions. In situ studies of pertinent electrochemical reactions in the molten salts are proposed, and are relevant to both the corrosive attack of structural materials, as well as the volatilization of fuel. Some consideration is given to the potential advantages of gravity fed falling-film blankets. Such systems may be easier to control than vortex systems, but would require that cylindrical reaction vessels be oriented with the centerline normal to the gravitational field.« less

  8. Integrated Risk-Informed Decision-Making for an ALMR PRISM

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

    Muhlheim, Michael David; Belles, Randy; Denning, Richard S.

    Decision-making is the process of identifying decision alternatives, assessing those alternatives based on predefined metrics, selecting an alternative (i.e., making a decision), and then implementing that alternative. The generation of decisions requires a structured, coherent process, or a decision-making process. The overall objective for this work is that the generalized framework is adopted into an autonomous decision-making framework and tailored to specific requirements for various applications. In this context, automation is the use of computing resources to make decisions and implement a structured decision-making process with limited or no human intervention. The overriding goal of automation is to replace ormore » supplement human decision makers with reconfigurable decision-making modules that can perform a given set of tasks rationally, consistently, and reliably. Risk-informed decision-making requires a probabilistic assessment of the likelihood of success given the status of the plant/systems and component health, and a deterministic assessment between plant operating parameters and reactor protection parameters to prevent unnecessary trips and challenges to plant safety systems. The probabilistic portion of the decision-making engine of the supervisory control system is based on the control actions associated with an ALMR PRISM. Newly incorporated into the probabilistic models are the prognostic/diagnostic models developed by Pacific Northwest National Laboratory. These allow decisions to incorporate the health of components into the decision–making process. Once the control options are identified and ranked based on the likelihood of success, the supervisory control system transmits the options to the deterministic portion of the platform. The deterministic portion of the decision-making engine uses thermal-hydraulic modeling and components for an advanced liquid-metal reactor Power Reactor Inherently Safe Module. The deterministic multi-attribute decision-making framework uses various sensor data (e.g., reactor outlet temperature, steam generator drum level) and calculates its position within the challenge state, its trajectory, and its margin within the controllable domain using utility functions to evaluate current and projected plant state space for different control decisions. The metrics that are evaluated are based on reactor trip set points. The integration of the deterministic calculations using multi-physics analyses and probabilistic safety calculations allows for the examination and quantification of margin recovery strategies. This also provides validation of the control options identified from the probabilistic assessment. Thus, the thermalhydraulics analyses are used to validate the control options identified from the probabilistic assessment. Future work includes evaluating other possible metrics and computational efficiencies, and developing a user interface to mimic display panels at a modern nuclear power plant.« less

  9. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    NASA Astrophysics Data System (ADS)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

  10. Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

    USGS Publications Warehouse

    Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

    2016-01-01

    Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

  11. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  12. The influence of pH adjustment on kinetics parameters in tapioca wastewater treatment using aerobic sequencing batch reactor system

    NASA Astrophysics Data System (ADS)

    Mulyani, Happy; Budianto, Gregorius Prima Indra; Margono, Kaavessina, Mujtahid

    2018-02-01

    The present investigation deals with the aerobic sequencing batch reactor system of tapioca wastewater treatment with varying pH influent conditions. This project was carried out to evaluate the effect of pH on kinetics parameters of system. It was done by operating aerobic sequencing batch reactor system during 8 hours in many tapioca wastewater conditions (pH 4.91, pH 7, pH 8). The Chemical Oxygen Demand (COD) and Mixed Liquor Volatile Suspended Solids (MLVSS) of the aerobic sequencing batch reactor system effluent at steady state condition were determined at interval time of two hours to generate data for substrate inhibition kinetics parameters. Values of the kinetics constants were determined using Monod and Andrews models. There was no inhibition constant (Ki) detected in all process variation of aerobic sequencing batch reactor system for tapioca wastewater treatment in this study. Furthermore, pH 8 was selected as the preferred aerobic sequencing batch reactor system condition in those ranging pH investigated due to its achievement of values of kinetics parameters such µmax = 0.010457/hour and Ks = 255.0664 mg/L COD.

  13. In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

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

    Burgett, Eric

    2015-10-13

    An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In additionmore » to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.« less

  14. 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.

  15. Component and System Sensitivity Considerations for Design of a Lunar ISRU Oxygen Production Plant

    NASA Technical Reports Server (NTRS)

    Linne, Diane L.; Gokoglu, Suleyman; Hegde, Uday G.; Balasubramaniam, Ramaswamy; Santiago-Maldonado, Edgardo

    2009-01-01

    Component and system sensitivities of some design parameters of ISRU system components are analyzed. The differences between terrestrial and lunar excavation are discussed, and a qualitative comparison of large and small excavators is started. The effect of excavator size on the size of the ISRU plant's regolith hoppers is presented. Optimum operating conditions of both hydrogen and carbothermal reduction reactors are explored using recently developed analytical models. Design parameters such as batch size, conversion fraction, and maximum particle size are considered for a hydrogen reduction reactor while batch size, conversion fraction, number of melt zones, and methane flow rate are considered for a carbothermal reduction reactor. For both reactor types the effect of reactor operation on system energy and regolith delivery requirements is presented.

  16. Microprocessor tester for the treat upgrade reactor trip system

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

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

  17. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  18. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

  19. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

  20. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

  1. Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

    NASA Technical Reports Server (NTRS)

    Gould, R. E.; Petticrew, R. W.

    1973-01-01

    This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

  2. Inhibiting mild steel corrosion from sulfate-reducing bacteria using antimicrobial-producing biofilms in Three-Mile-Island process water.

    PubMed

    Zuo, R; Ornek, D; Syrett, B C; Green, R M; Hsu, C-H; Mansfeld, F B; Wood, T K

    2004-04-01

    Biofilms were used to produce gramicidin S (a cyclic decapeptide) to inhibit corrosion-causing, sulfate-reducing bacteria (SRB). In laboratory studies these biofilms protected mild steel 1010 continuously from corrosion in the aggressive, cooling service water of the AmerGen Three-Mile-Island (TMI) nuclear plant, which was augmented with reference SRB. The growth of both reference SRB (Gram-positive Desulfosporosinus orientis and Gram-negative Desulfovibrio vulgaris) was shown to be inhibited by supernatants of the gramicidin-S-producing bacteria as well as by purified gramicidin S. Electrochemical impedance spectroscopy and mass loss measurements showed that the protective biofilms decreased the corrosion rate of mild steel by 2- to 10-fold when challenged with the natural SRB of the TMI process water supplemented with D. orientis or D. vulgaris. The relative corrosion inhibition efficiency was 50-90% in continuous reactors, compared to a biofilm control which did not produce the antimicrobial gramicidin S. Scanning electron microscope and reactor images also revealed that SRB attack was thwarted by protective biofilms that secrete gramicidin S. A consortium of beneficial bacteria (GGPST consortium, producing gramicidin S and other antimicrobials) also protected the mild steel.

  3. Demonstration of Robustness and Integrated Operation of a Series-Bosch System

    NASA Technical Reports Server (NTRS)

    Abney, Morgan B.; Mansell, J. Matthew; Barnett, Bill; Stanley, Christine M.; Junaedi, Christian; Vilekar, Saurabh A.; Kent, Ryan

    2016-01-01

    Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a RWGS reactor containing Incofoam(TradeMark) catalyst and designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith(TradeMark) technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The Microlith(TradeMark) RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with the Incofoam(TradeMark) RWGS reactor. Separately, in 2015, a fully integrated demonstration of an S-Bosch system was conducted. In an effort to mitigate risk, a second integrated test was conducted to evaluate the effect of membrane failure on a closed-loop Bosch system. Here, we report and discuss the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level. 1

  4. Annual report to Congress, FY 1992

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

    NONE

    1993-07-01

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for disposing of the Nation`s spent nuclear fuel from civilian nuclear power reactors and high-level radioactive waste from its defense activities in a cost-effective manner that protects the health and safety of the public and workers and the quality of the environment. To accomplish this mission OCRWM is developing a waste management system consisting of a geologic repository, a facility for monitored retrievable storage, and a system for transporting the waste. This is the ninth annual report submitted by the OCRWM to Congress. The OCRWM submits this report to informmore » Congress of its activities and expenditures during fiscal year 1992 (October 1, 1991 through September 30, 1992).« less

  5. The development of regulatory expectations for computer-based safety systems for the UK nuclear programme

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

    Hughes, P. J.; Westwood, R.N; Mark, R. T.

    2006-07-01

    The Nuclear Installations Inspectorate (NII) of the UK's Health and Safety Executive (HSE) has completed a review of their Safety Assessment Principles (SAPs) for Nuclear Installations recently. During the period of the SAPs review in 2004-2005 the designers of future UK naval reactor plant were optioneering the control and protection systems that might be implemented. Because there was insufficient regulatory guidance available in the naval sector to support this activity the Defence Nuclear Safety Regulator (DNSR) invited the NII to collaborate with the production of a guidance document that provides clarity of regulatory expectations for the production of safety casesmore » for computer based safety systems. A key part of producing regulatory expectations was identifying the relevant extant standards and sector guidance that reflect good practice. The three principal sources of such good practice were: IAEA Safety Guide NS-G-1.1 (Software for Computer Based Systems Important to Safety in Nuclear Power Plants), European Commission consensus document (Common Position of European Nuclear Regulators for the Licensing of Safety Critical Software for Nuclear Reactors) and IEC nuclear sector standards such as IEC60880. A common understanding has been achieved between the NII and DNSR and regulatory guidance developed which will be used by both NII and DNSR in the assessment of computer-based safety systems and in the further development of more detailed joint technical assessment guidance for both regulatory organisations. (authors)« less

  6. On the radiation damage characterization of candidate first wall materials in a fusion reactor using various molten salts

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa

    2006-12-01

    Evaluating radiation damage characteristics of structural materials considered to be used in fusion reactors is very crucial. In fusion reactors, the highest material damage occurs in the first wall because it will be exposed to the highest neutron, gamma ray and charged particle currents produced in the fusion chamber. This damage reduces the lifetime of the first wall material and leads to frequent replacement of this material during the reactor operation period. In order to decrease operational cost of a fusion reactor, lifetime of the first wall material should be extended to reactor's lifetime. Using a protective flowing liquid wall between the plasma and first wall can decrease the radiation damage on first wall and extend its lifetime to the reactor's lifetime. In this study, radiation damage characterization of various low activation materials used as first wall material in a magnetic fusion reactor blanket using a liquid wall was made. Various coolants (Flibe, Flibe + 4% mol ThF 4, Flibe + 8% mol ThF 4, Li 20Sn 80) were used to investigate their effect on the radiation damage of first wall materials. Calculations were carried out by using the code Scale4.3 to solve Boltzmann neutron transport equation. Numerical results brought out that the ferritic steel with Flibe based coolants showed the best performance with respect to radiation damage.

  7. JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

    Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

    1963-01-01

    ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

  8. Method of producing gaseous products using a downflow reactor

    DOEpatents

    Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

    2014-09-16

    Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

  9. RADON LEVELS AND ЕQUIVALENT DOSE RATES AT THE IRT-SOFIA RESEARCH REACTOR SITE.

    PubMed

    Krezhov, Kiril; Mladenov, Aleksander; Dimitrov, Dobromir

    2018-06-11

    Results from radon measurements by active sampling of indoor air in the buildings within the Nuclear Scientific Experimental and Educational Centre (NSEEC) protected site at the Institute for Nuclear Research and Nuclear Energy (INRNE) are presented. The inspected buildings included in this report are the IRT research reactor structure and several auxiliary formations wherein the laundry facilities and the gamma irradiator GOU-1 (60Co source) are installed as well as the Central Alarm Station (CAS) premises. Besides the reactor hall and the primary cooling loop area, special attention was given to the premises of the First Class Radiochemical Laboratory in the IRT reactor basement. Determination of radon concentration distribution in the premises of the constructions within the site is an important part of radiation surveillance during the operation and maintenance of the NSEEC facilities as well as for their involvement in the educational activities at INRNE.

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

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

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

  11. 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.

  12. 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.

  13. Co-Production of Electricity and Hydrogen Using a Novel Iron-based Catalyst

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

    Hilaly, Ahmad; Georgas, Adam; Leboreiro, Jose

    2011-09-30

    The primary objective of this project was to develop a hydrogen production technology for gasification applications based on a circulating fluid-bed reactor and an attrition resistant iron catalyst. The work towards achieving this objective consisted of three key activities: Development of an iron-based catalyst suitable for a circulating fluid-bed reactor; Design, construction, and operation of a bench-scale circulating fluid-bed reactor system for hydrogen production; Techno-economic analysis of the steam-iron and the pressure swing adsorption hydrogen production processes. This report describes the work completed in each of these activities during this project. The catalyst development and testing program prepared and iron-basedmore » catalysts using different support and promoters to identify catalysts that had sufficient activity for cyclic reduction with syngas and steam oxidation and attrition resistance to enable use in a circulating fluid-bed reactor system. The best performing catalyst from this catalyst development program was produced by a commercial catalyst toll manufacturer to support the bench-scale testing activities. The reactor testing systems used during material development evaluated catalysts in a single fluid-bed reactor by cycling between reduction with syngas and oxidation with steam. The prototype SIP reactor system (PSRS) consisted of two circulating fluid-bed reactors with the iron catalyst being transferred between the two reactors. This design enabled demonstration of the technical feasibility of the combination of the circulating fluid-bed reactor system and the iron-based catalyst for commercial hydrogen production. The specific activities associated with this bench-scale circulating fluid-bed reactor systems that were completed in this project included design, construction, commissioning, and operation. The experimental portion of this project focused on technical demonstration of the performance of an iron-based catalyst and a circulating fluid-bed reactor system for hydrogen production. Although a technology can be technically feasible, successful commercial deployment also requires that a technology offer an economic advantage over existing commercial technologies. To effective estimate the economics of this steam-iron process, a techno-economic analysis of this steam iron process and a commercial pressure swing adsorption process were completed. The results from this analysis described in this report show the economic potential of the steam iron process for integration with a gasification plant for coproduction of hydrogen and electricity.« less

  14. Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

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

    Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

    The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

  15. Production assurance program strategy for N Reactor balance of plant systems

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

    House, R.D.; Bitten, E.J.; Keenan, J.P.

    1986-03-18

    A production assurance program has been established for N Reactor, a dual purpose reactor plant, operated to produce special nuclear materials and steam for electricity. N Reactor, which began operation in December 1963, is now approaching the end of its design life. This paper describes the two phase program for Balance of Plant (BOP) systems. The Phase I evaluation has been completed and indications are that the lifetime of systems and components could be extended by implementing appropriate surveillance, operations and maintenance strategies. In Phase II, a thorough evaluation of components and systems is underway and action items are beingmore » identified which will allow component and system extended operation.« less

  16. Manual of Protective Action Guides and Protective Actions for Nuclear Incidents. Revision

    DTIC Science & Technology

    1980-06-01

    acceptable dose. Simce the PAC In based on a projected does, It is used only In an expost facto effort to minmisea the risk from an swest which is occurring...statistical evaluation of epidmiological studies In groups of people wbo had been ezposed to radiation. Decisions concerning statistical effects on...protective actions. The Reactor Saety Study Indicates, for ezample, that major releases y beoon in the range of one-half how to as uab as 30 hours after an

  17. Small Reactor for Deep Space Exploration

    ScienceCinema

    none,

    2018-06-06

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  18. 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.

  19. PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

    Murray, J.L.

    1961-02-01

    BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

  20. System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

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

    Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

    2002-07-01

    Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

  1. System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

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

    Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

    2004-03-15

    Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

  2. 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.

  3. CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

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

    Ligeti, G.

    1962-04-01

    Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

  4. Inherently Safe Fission Power System for Lunar Outposts

    NASA Astrophysics Data System (ADS)

    Schriener, Timothy M.; El-Genk, Mohamed S.

    2013-09-01

    This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

  5. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOEpatents

    Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

    1992-01-01

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  6. Process for removing copper in a recoverable form from solid scrap metal

    DOEpatents

    Hartman, Alan D.; Oden, Laurance L.; White, Jack C.

    1995-01-01

    A process for removing copper in a recoverable form from a copper/solid ferrous scrap metal mix is disclosed. The process begins by placing a copper/solid ferrous scrap metal mix into a reactor vessel. The atmosphere within the reactor vessel is purged with an inert gas or oxidizing while the reactor vessel is heated in the area of the copper/solid ferrous scrap metal mix to raise the temperature within the reactor vessel to a selected elevated temperature. Air is introduced into the reactor vessel and thereafter hydrogen chloride is introduced into the reactor vessel to obtain a desired air-hydrogen chloride mix. The air-hydrogen chloride mix is operable to form an oxidizing and chloridizing atmosphere which provides a protective oxide coating on the surface of the solid ferrous scrap metal in the mix and simultaneously oxidizes/chloridizes the copper in the mix to convert the copper to a copper monochloride gas for transport away from the solid ferrous scrap metal. After the copper is completely removed from the copper/solid ferrous scrap metal mix, the flows of air and hydrogen chloride are stopped and the copper monochloride gas is collected for conversion to a recoverable copper species.

  7. 76 FR 55718 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-09-08

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor...'' for reactor coolant system (RCS) components, as mentioned in 10 CFR 50 Appendix A, GDC-4. The...

  8. 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.

  9. Assessing pretreatment reactor scaling through empirical analysis

    DOE PAGES

    Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; ...

    2016-10-10

    Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was within the near-optimal space for total sugar yield for the LHR. This indicates that the ASE is a good tool for cost effectively finding near-optimal conditions for operating pilot-scale systems, which may be used as starting points for further optimization. Additionally, using a severity-factor approach to optimization was found to be inadequate compared to a multivariate optimization method. As a result, the ASE and the LHR were able to enable significantly higher total sugar yields after enzymatic hydrolysis relative to the ZCR, despite having similar optimal conditions and total xylose yields. This underscores the importance of incorporating mechanical disruption into pretreatment reactor designs to achieve high enzymatic digestibilities.« less

  10. Assessing pretreatment reactor scaling through empirical analysis

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

    Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik

    Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, thismore » is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was within the near-optimal space for total sugar yield for the LHR. This indicates that the ASE is a good tool for cost effectively finding near-optimal conditions for operating pilot-scale systems, which may be used as starting points for further optimization. Additionally, using a severity-factor approach to optimization was found to be inadequate compared to a multivariate optimization method. As a result, the ASE and the LHR were able to enable significantly higher total sugar yields after enzymatic hydrolysis relative to the ZCR, despite having similar optimal conditions and total xylose yields. This underscores the importance of incorporating mechanical disruption into pretreatment reactor designs to achieve high enzymatic digestibilities.« less

  11. When Do Commercial Reactors Permanently Shut Down?

    EIA Publications

    2011-01-01

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

  12. Materials corrosion and protection from first principles

    NASA Astrophysics Data System (ADS)

    Johnson, Donald F.

    Materials erode under environmental stresses such as high temperature, high pressure, and mechanical shock/stress, but erosion is often exacerbated by chemical corrosion. In this dissertation, periodic density functional theory (DFT) is employed to simulate interfacial adhesion, absorption kinetics, bulk diffusion, and other material phenomena (e.g., hydrogen-enhanced decohesion and shock-induced phase changes) with the intention of understanding corrosion and subsequent failure processes and guiding the design of new protective coatings. This work examines corrosion and/or protection of materials ( i.e., Fe, Ni, W) with important applications: structural steel, gun tubes, high-pressure oil recovery vessels, jet engine turbine blades, and fusion reactor walls. We use DFT to model the pressure-induced, bcc-to-hcp phase transformation in Fe, in which a new low energy pathway is predicted exhibiting nonadiabatic behavior coupling magnetic and structural changes. Protection of steel is addressed in two aspects: interfacial adhesion of protective coatings and assessment of corrosion resistance provided by a surface alloy. First, the current chrome-coated steel system is examined where extremely strong adhesion is predicted at the Cr/Fe interface originating in strong spin correlations. A ceramic coating, SiC, is considered as a possible replacement for Cr. Strong adhesion is predicted, especially for C-Fe interfacial bonds. To assess corrosion resistance, we model ingress of two common corrosive elements, H and C, into two Fe alloys, FeAl and Fe3Si. Adsorption and absorption thermodynamics and kinetics, as well as bulk dissolution and diffusion are calculated in order to determine whether these two alloys can inhibit uptake of H and C. Relative to pure Fe, dissolved H and C are less stable in the alloys, as the dissolution enthalpy is predicted to be more endothermic. Overall, the energy barriers and rate constants for adsorbed H/C diffusing into Fe3Si subsurface layers suggests that alloying Fe with Si can be an effective means to limit uptake of these elements into steel. Spallation of protective layers on jet engine turbine blades is a problem that arises during thermal cycling. An alternative thermal barrier coating system involving MoSi2 is considered and calculations predict strong adhesion at the MoSi2/Ni interface. The interfacial bonding structure reveals a mixture of metallic and covalent cross-interface bonds. The adhesion energy is similar across all three MoSi2 facets studied. Upon exposure to oxygen, this MoSi2 alloy will form a strongly adhered oxide scale, which in turn may strongly adhere the heat shield material (yttria-stabilized zirconia), thereby potentially extending the lifetime of the barrier coating. Lastly, the interaction of hydrogen isotopes (fusion fuel) with tungsten (a proposed fusion reactor wall material) is examined. Exothermic dissociative adsorption is predicted, along with endothermic absorption and dissolution. Surface-to-subsurface diffusion energy barriers for H incorporation into bulk W are large and the corresponding outward diffusion barriers are very small. In bulk W, deep energetic traps (trapping multiple H atoms) are predicted at vacancy defects. Thus, under high neutron fluxes that will produce vacancies in W, H are predicted to collect at these vacancies. In turn, locally high concentrations of H at such vacancies will enhance decohesion of bulk W, consistent with observed blistering under deuterium implantation. Limiting vacancy formation may be key to the survival of W as a fusion reactor wall material.

  13. Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

    NASA Astrophysics Data System (ADS)

    Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

    2010-10-01

    The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

  14. Carbon monoxide oxidation rates computed for automobile thermal reactor conditions

    NASA Technical Reports Server (NTRS)

    Brokaw, R. S.; Bittker, D. A.

    1972-01-01

    Carbon monoxide oxidation rates in thermal reactors for exhaust manifolds are computed by integrating differential equations for system of twenty-nine reversible chemical reactions. Reactors are noncatalytic replacements for conventional exhaust manifolds and are a system for reducing carbon monoxide and hydrocarbons in automobile exhausts.

  15. Modeling of a Reaction-Distillation-Recycle System to Produce Dimethyl Ether through Methanol Dehydration

    NASA Astrophysics Data System (ADS)

    Muharam, Y.; Zulkarnain, L. M.; Wirya, A. S.

    2018-03-01

    The increase in the dimethyl ether yield through methanol dehydration due to a recycle integration to a reaction-distillation system was studied in this research. A one-dimensional phenomenological model of a methanol dehydration reactor and a shortcut model of distillation columns were used to achieve the aim. Simulation results show that 10.7 moles/s of dimethyl ether is produced in a reaction-distillation system with the reactor length being 4 m, the reactor inlet pressure being 18 atm, the reactor inlet temperature being 533 K, the reactor inlet velocity being 0.408 m/s, and the distillation pressure being 8 atm. The methanol conversion is 90% and the dimethyl ether yield is 48%. The integration of the recycle stream to the system increases the dimethyl ether yield by 8%.

  16. 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.

  17. 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.

  18. 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.

  19. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-19

    ... 100 Grams,'' dated October 10, 2002, and subsequent similar security orders issued to licensees... than 100 grams from theft, diversion, or radiological sabotage. The requirements in the final rule...

  20. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

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

    Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.

    2016-01-01

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologiesmore » for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.« less

  1. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    DOE PAGES

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; ...

    2017-01-24

    We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

  2. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  3. Autonomous Control of Nuclear Power Plants

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

    Basher, H.

    2003-10-20

    A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that maymore » be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.« less

  4. 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.

  5. Catalytic Reactor for Inerting of Aircraft Fuel Tanks

    DTIC Science & Technology

    1974-06-01

    Aluminum Panels After Triphase Corrosion Test 79 35 Inerting System Flows in Various Flight Modes 82 36 High Flow Reactor Parametric Data 84 37 System...AD/A-000 939 CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS George H. McDonald, et al AiResearch Manufacturing Company Prepared for: Air Force...190th Street 2b. GROUP Torrance, California .. REPORT TITLE CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS . OESCRIP TIVE NOTEs (Thpe of refpoft

  6. Fission control system for nuclear reactor

    DOEpatents

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  7. Statistical Exposé of a Multiple-Compartment Anaerobic Reactor Treating Domestic Wastewater.

    PubMed

    Pfluger, Andrew R; Hahn, Martha J; Hering, Amanda S; Munakata-Marr, Junko; Figueroa, Linda

    2018-06-01

      Mainstream anaerobic treatment of domestic wastewater is a promising energy-generating treatment strategy; however, such reactors operated in colder regions are not well characterized. Performance data from a pilot-scale, multiple-compartment anaerobic reactor taken over 786 days were subjected to comprehensive statistical analyses. Results suggest that chemical oxygen demand (COD) was a poor proxy for organics in anaerobic systems as oxygen demand from dissolved inorganic material, dissolved methane, and colloidal material influence dissolved and particulate COD measurements. Additionally, univariate and functional boxplots were useful in visualizing variability in contaminant concentrations and identifying statistical outliers. Further, significantly different dissolved organic removal and methane production was observed between operational years, suggesting that anaerobic reactor systems may not achieve steady-state performance within one year. Last, modeling multiple-compartment reactor systems will require data collected over at least two years to capture seasonal variations of the major anaerobic microbial functions occurring within each reactor compartment.

  8. Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

    PubMed

    Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

    2016-04-01

    A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. Employing ISRU Models to Improve Hardware Design

    NASA Technical Reports Server (NTRS)

    Linne, Diane L.

    2010-01-01

    An analytical model for hydrogen reduction of regolith was used to investigate the effects of several key variables on the energy and mass performance of reactors for a lunar in-situ resource utilization oxygen production plant. Reactor geometry, reaction time, number of reactors, heat recuperation, heat loss, and operating pressure were all studied to guide hardware designers who are developing future prototype reactors. The effects of heat recuperation where the incoming regolith is pre-heated by the hot spent regolith before transfer was also investigated for the first time. In general, longer reaction times per batch provide a lower overall energy, but also result in larger and heavier reactors. Three reactors with long heat-up times results in similar energy requirements as a two-reactor system with all other parameters the same. Three reactors with heat recuperation results in energy reductions of 20 to 40 percent compared to a three-reactor system with no heat recuperation. Increasing operating pressure can provide similar energy reductions as heat recuperation for the same reaction times.

  10. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

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

  11. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    NASA Astrophysics Data System (ADS)

    Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

    2004-02-01

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  12. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

    2005-02-01

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  13. DEVELOPMENT, INSTALLATION AND OPERATION OF THE MPC&A OPERATIONS MONITORING (MOM) SYSTEM AT THE JOINT INSTITUTE FOR NUCLEAR RESEARCH (JINR) DUBNA, RUSSIA

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

    Kartashov,V.V.; Pratt,W.; Romanov, Y.A.

    The Material Protection, Control and Accounting (MPC&A) Operations Monitoring (MOM) systems handling at the International Intergovernmental Organization - Joint Institute for Nuclear Research (JINR) is described in this paper. Category I nuclear material (plutonium and uranium) is used in JINR research reactors, facilities and for scientific and research activities. A monitoring system (MOM) was installed at JINR in April 2003. The system design was based on a vulnerability analysis, which took into account the specifics of the Institute. The design and installation of the MOM system was a collaborative effort between JINR, Brookhaven National Laboratory (BNL) and the U.S. Departmentmore » of Energy (DOE). Financial support was provided by DOE through BNL. The installed MOM system provides facility management with additional assurance that operations involving nuclear material (NM) are correctly followed by the facility personnel. The MOM system also provides additional confidence that the MPC&A systems continue to perform effectively.« less

  14. 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.

  15. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  16. The components of shear stress affecting insect cells used with the baculovirus expression vector system.

    PubMed

    Weidner, Tobias; Druzinec, Damir; Mühlmann, Martina; Buchholz, Rainer; Czermak, Peter

    2017-09-26

    Insect-based expression platforms such as the baculovirus expression vector system (BEVS) are widely used for the laboratory- and industrial-scale production of recombinant proteins. Thereby, major drawbacks to gain high-quality proteins are the lytic infection cycle and the shear sensitivity of infected insect cells due to turbulence and aeration. Smaller bubbles were formerly assumed to be more harmful than larger ones, but we found that cell damage is also dependent on the concentration of protective agents such as Pluronic®. At the appropriate concentration, Pluronic forms a layer around air bubbles and hinders the attachment of cells, thus limiting the damage. In this context, we used microaeration to vary bubble sizes and confirmed that size is not the most important factor, but the total gas surface area in the reactor is. If the surface area exceeds a certain threshold, the concentration of Pluronic is no longer sufficient for cell protection. To investigate the significance of shear forces, a second study was carried out in which infected insect cells were cultivated in a hollow fiber module to protect them from shear forces. Both model studies revealed important aspects of the design and scale-up of BEVS processes for the production of recombinant proteins.

  17. World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

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

    Not Available

    1979-06-01

    Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

  18. A study on using fireclay as a biomass carrier in an activated sludge system.

    PubMed

    Tilaki, Ramazan Ali Dianati

    2011-01-01

    By adding a biomass carrier to an activated sludge system, the biomass concentration will increase, and subsequently the organic removal efficiency will be enhanced. In this study, the possibility of using excess sludge from ceramic and tile manufacturing plants as a biomass carrier was investigated. The aim of this study was to determine the effect of using fireclay as a biomass carrier on biomass concentration, organic removal and nitrification efficiency in an activated sludge system. Experiments were conducted by using a bench scale activated sludge system operating in batch and continuous modes. Artificial simulated wastewater was made by using recirculated water in a ceramic manufacturing plant. In the continuous mode, hydraulic detention time in the aeration reactor was 8 and 22 h. In the batch mode, aeration time was 8 and 16 h. Fireclay doses were 500, 1,400 and 2,250 mg l(-1), and were added to the reactors in each experiment separately. The reactor with added fireclay was called a Hybrid Biological Reactor (HBR). A reactor without added fireclay was used as a control. Efficiency parameters such as COD, MLVSS and nitrate were measured in the control and HBR reactors according to standard methods. The average concentration of biomass in the HBR reactor was greater than in the control reactor. The total biomass concentration in the HBR reactor (2.25 g l(-1) fireclay) in the continuous mode was 3,000 mg l(-1) and in the batch mode was 2,400 mg l(-1). The attached biomass concentration in the HBR reactor (2.25 g l(-1) fireclay) in the continuous mode was 1,500 mg l(-1) and in the batch mode was 980 mg l(-1). Efficiency for COD removal in the HBR and control reactor was 95 and 55%, respectively. In the HBR reactor, nitrification was enhanced, and the concentration of nitrate was increased by 80%. By increasing the fireclay dose, total and attached biomass was increased. By adding fireclay as a biomass carrier, the efficiency of an activated sludge system to treat wastewater from ceramic manufacturing plants was increased.

  19. Modeling and simulation of enzymatic gluconic acid production using immobilized enzyme and CSTR-PFTR circulation reaction system.

    PubMed

    Li, Can; Lin, Jianqun; Gao, Ling; Lin, Huibin; Lin, Jianqiang

    2018-04-01

    Production of gluconic acid by using immobilized enzyme and continuous stirred tank reactor-plug flow tubular reactor (CSTR-PFTR) circulation reaction system. A production system is constructed for gluconic acid production, which consists of a continuous stirred tank reactor (CSTR) for pH control and liquid storage and a plug flow tubular reactor (PFTR) filled with immobilized glucose oxidase (GOD) for gluconic acid production. Mathematical model is developed for this production system and simulation is made for the enzymatic reaction process. The pH inhibition effect on GOD is modeled by using a bell-type curve. Gluconic acid can be efficiently produced by using the reaction system and the mathematical model developed for this system can simulate and predict the process well.

  20. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

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