Some Implications of the Three Mile Island Accident for LMFBR Safety and Licensing: The Design Basis Issue

DTIC Science & Technology

1980-08-01

metal fast breeder reactor (LMFBR) design. It also re-examines the impact of the accident at Three Mile Island on the design basis concept, and how...Water Reactors : ImpZications for Liquid MetaZ Fast Breeder Reactors , by W. E. Kastenberg and K. A. Solomon, July 1979. v SUNMARY The 1979 accident...the liquid metal fast breeder reactor (LMFBR). This Note assesses the impact of the TMI-2 accident on the LMFBR. Specifically, it: o Reviews the

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

10 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...

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...

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...

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...

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

Pind, C.

The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives formore » heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories.« less

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

DTIC Science & Technology

2008-11-03

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

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

DTIC Science & Technology

2008-10-02

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

Nuclear Security: Action May Be Needed to Reassess the Security of NRC-Licensed Research Reactors. Report to the Ranking Member, Subcommittee on National Security and Foreign Affairs, Committee on Oversight and Government Reform, House of Representatives. GAO-08-403

ERIC Educational Resources Information Center

Aloise, Gene

2008-01-01

There are 37 research reactors in the United States, mostly located on college campuses. Of these, 33 reactors are licensed and regulated by the Nuclear Regulatory Commission (NRC). Four are operated by the Department of Energy (DOE) and are located at three national laboratories. Although less powerful than commercial nuclear power reactors,…

Eddy Current Flow Measurements in the FFTF

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

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

Benchmark tests of JENDL-3.2 for thermal and fast reactors

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

Takano, Hideki; Akie, Hiroshi; Kikuchi, Yasuyuki

1994-12-31

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k{sub eff} and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k{sub eff} reactivity worths of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments.

Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

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

Takeda, T.; Shimazu, Y.; Hibi, K.

2012-07-01

Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of thismore » project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)« less

Simulator platform for fast reactor operation and safety technology demonstration

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

Vilim, R. B.; Park, Y. S.; Grandy, C.

2012-07-30

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe responsemore » to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.« less

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

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

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

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

2012-11-01

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

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

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

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

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

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

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

The benefits of a fast reactor closed fuel cycle in the UK

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

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

Japan’s Nuclear Future: Policy Debate, Prospects, and U.S. Interests

DTIC Science & Technology

2008-05-09

raised in particular over the construction of an industrial- scale reprocessing facility in Japan,. Additionally, fast breeder reactors also produce more...Nuclear Fuel Cycle Engineering Laboratories. 10 A fast breeder reactor is a fast neutron reactor that produces more plutonium than it consumes, which can...Japan Nuclear Fuel Limited (JNFL) has built and is currently running active testing on a large - scale commercial reprocessing plant at Rokkasho-mura

Significance of breeding in fast nuclear reactors

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

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

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

Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

2010-01-01

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.« less

Thermomechanical analysis of fast-burst reactors

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

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Fast quench reactor and method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

1998-05-12

A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

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

DTIC Science & Technology

2008-10-17

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

Conceptual design of laser fusion reactor KOYO-fast Concepts of reactor system and laser driver

NASA Astrophysics Data System (ADS)

Kozaki, Y.; Miyanaga, N.; Norimatsu, T.; Soman, Y.; Hayashi, T.; Furukawa, H.; Nakatsuka, M.; Yoshida, K.; Nakano, H.; Kubomura, H.; Kawashima, T.; Nishimae, J.; Suzuki, Y.; Tsuchiya, N.; Kanabe, T.; Jitsuno, T.; Fujita, H.; Kawanaka, J.; Tsubakimoto, K.; Fujimoto, Y.; Lu, J.; Matsuoka, S.; Ikegawa, T.; Owadano, Y.; Ueda, K.; Tomabechi, K.; Reactor Design Committee in Ife Forum, Members Of

2006-06-01

We have carried out the design studies of KOYO-Fast laser fusion power plant, using fast ignition cone targets, DPSSL lasers, and LiPb liquid wall chambers. Using fast ignition targets, we could design a middle sized 300 MWe reactor module, with 200 MJ fusion pulse energy and 4 Hz rep-rates, and 1200MWe modular power plants with 4 reactor modules and a 16 Hz laser driver. The liquid wall chambers with free surface cascade flows are proposed for cooling surface quickly enough to a 4 Hz pulse operation. We examined the potential of Yb-YAG ceramic lasers operated at 150˜ 225 K for both implosion and heating laser systems required for a 16-Hz repetition and 8 % total efficiency.

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

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

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

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

Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uraniummore » fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.« less

FAST CHOPPER BUILDING, TRA665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. ...

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

FAST CHOPPER BUILDING, TRA-665. CONTEXTUAL VIEW: CHOPPER BUILDING IN CENTER. MTR REACTOR SERVICES BUILDING,TRA-635, TO LEFT; MTR BUILDING TO RIGHT. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

COUPLED FAST-THERMAL POWER BREEDER REACTOR

DOEpatents

Avery, R.

1961-07-18

A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

MTR STACK, TRA710, CONTEXTUAL VIEW, CAMERA FACING SOUTH. PERIMETER SECURITY ...

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

MTR STACK, TRA-710, CONTEXTUAL VIEW, CAMERA FACING SOUTH. PERIMETER SECURITY FENCE AND SECURITY LIGHTING IN VIEW AT LEFT. INL NEGATIVE NO. HD52-1-1. Mike Crane, Photographer, 5/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Biodiesel production process intensification using a rotor-stator type generator of hydrodynamic cavitation.

PubMed

Crudo, Daniele; Bosco, Valentina; Cavaglià, Giuliano; Grillo, Giorgio; Mantegna, Stefano; Cravotto, Giancarlo

2016-11-01

Triglyceride transesterification for biodiesel production is a model reaction which is used to compare the conversion efficiency, yield, reaction time, energy consumption, scalability and cost estimation of different reactor technology and energy source. This work describes an efficient, fast and cost-effective procedure for biodiesel preparation using a rotating generator of hydrodynamic cavitation (HC). The base-catalyzed transesterification (methanol/sodium hydroxide) has been carried out using refined and bleached palm oil and waste vegetable cooking oil. The novel HC unit is a continuous rotor-stator type reactor in which reagents are directly fed into the controlled cavitation chamber. The high-speed rotation of the reactor creates micron-sized droplets of the immiscible reacting mixture leading to outstanding mass and heat transfer and enhancing the kinetics of the transesterification reaction which completes much more quickly than traditional methods. All the biodiesel samples obtained respect the ASTM standard and present fatty acid methyl ester contents of >99% m/m in both feedstocks. The electrical energy consumption of the HC reactor is 0.030kWh per L of produced crude biodiesel, making this innovative technology really quite competitive. The reactor can be easily scaled-up, from producing a few hundred to thousands of liters of biodiesel per hour while avoiding the risk of orifices clogging with oil impurities, which may occur in conventional HC reactors. Furthermore it requires minimal installation space due to its compact design, which enhances overall security. Copyright © 2016 Elsevier B.V. All rights reserved.

75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-24

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0228] Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents...

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

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

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« less

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

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

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

2016-09-16

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

Development of Cross Section Library and Application Programming Interface (API)

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

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

Homogeneous fast-flux isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

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

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

77 FR 49048 - Self-Regulatory Organizations; The Depository Trust Company; Notice of Filing and Immediate...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-15

... to certain issues of securities in DTC's Fast Automated Securities Transfer program (``FAST''). For securities that are part of the FAST program, the transfer agents hold the securities registered in the name...

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

Radioactive waste from decommissioning of fast reactors (through the example of BN-800)

NASA Astrophysics Data System (ADS)

Rybin, A. A.; Momot, O. A.

2017-01-01

Estimation of volume of radioactive waste from operating and decommissioning of fast reactors is introduced. Preliminary estimation has shown that the volume of RW from decommissioning of BN-800 is amounted to 63,000 cu. m. Comparison of the amount of liquid radioactive waste derived from operation of different reactor types is performed. Approximate costs of all wastes disposal for complete decommissioning of BN-800 reactor are estimated amounting up to approx. 145 million.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

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

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

Not Available

1992-07-01

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

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

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

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

1998-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-09-24

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Performance of compact fast pyrolysis reactor with Auger-type modules for the continuous liquid biofuel production

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

Development of a compact fast pyrolysis reactor constructed using Auger-type technology to afford liquid biofuel with high yield has been an interesting concept in support of local production for local consumption. To establish a widely useable module package, details of the performance of the developing compact module reactor were investigated. This study surveyed the properties of as-produced pyrolysis oil as a function of operation time, and clarified the recent performance of the developing compact fast pyrolysis reactor. Results show that after condensation in the scrubber collector, e.g. approx. 10 h for a 25 kg/h feedstock rate, static performance of pyrolysis oil with approximately 20 MJ/kg (4.8 kcal/g) calorific values were constantly obtained after an additional 14 h. The feeding speed of cedar chips strongly influenced the time for oil condensation process: i.e. 1.6 times higher feeding speed decreased the condensation period by half (approx. 5 h in the case of 40 kg/h). Increasing the reactor throughput capacity is an important goal for the next stage in the development of a compact fast pyrolysis reactor with Auger-type modules.

One-dimensional biomass fast pyrolysis model with reaction kinetics integrated in an Aspen Plus Biorefinery Process Model

DOE PAGES

Humbird, David; Trendewicz, Anna; Braun, Robert; ...

2017-01-12

A biomass fast pyrolysis reactor model with detailed reaction kinetics and one-dimensional fluid dynamics was implemented in an equation-oriented modeling environment (Aspen Custom Modeler). Portions of this work were detailed in previous publications; further modifications have been made here to improve stability and reduce execution time of the model to make it compatible for use in large process flowsheets. The detailed reactor model was integrated into a larger process simulation in Aspen Plus and was stable for different feedstocks over a range of reactor temperatures. Sample results are presented that indicate general agreement with experimental results, but with higher gasmore » losses caused by stripping of the bio-oil by the fluidizing gas in the simulated absorber/condenser. Lastly, this integrated modeling approach can be extended to other well-defined, predictive reactor models for fast pyrolysis, catalytic fast pyrolysis, as well as other processes.« less

One-dimensional biomass fast pyrolysis model with reaction kinetics integrated in an Aspen Plus Biorefinery Process Model

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

Humbird, David; Trendewicz, Anna; Braun, Robert

A biomass fast pyrolysis reactor model with detailed reaction kinetics and one-dimensional fluid dynamics was implemented in an equation-oriented modeling environment (Aspen Custom Modeler). Portions of this work were detailed in previous publications; further modifications have been made here to improve stability and reduce execution time of the model to make it compatible for use in large process flowsheets. The detailed reactor model was integrated into a larger process simulation in Aspen Plus and was stable for different feedstocks over a range of reactor temperatures. Sample results are presented that indicate general agreement with experimental results, but with higher gasmore » losses caused by stripping of the bio-oil by the fluidizing gas in the simulated absorber/condenser. Lastly, this integrated modeling approach can be extended to other well-defined, predictive reactor models for fast pyrolysis, catalytic fast pyrolysis, as well as other processes.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

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

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

FAST CHOPPER BUILDING, TRA665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL ...

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

FAST CHOPPER BUILDING, TRA-665, INTERIOR. UPPER LEVEL. CONCRETE WALLS. INL NEGATIVE NO. HD42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

A Comparison of Fast-Spectrum and Moderated Space Fission Reactors

NASA Astrophysics Data System (ADS)

Poston, David I.

2005-02-01

The reactor neutron spectrum is one of the fundamental design choices for any fission reactor, but the implications of using a moderated spectrum are vastly different for space reactors as opposed to terrestrial reactors. In addition, the pros and cons of neutron spectra are significantly different among many of the envisioned space power applications. This paper begins with a discussion of the neutronic differences between fast-spectrum and moderated space reactors. This is followed by a discussion of the pros and cons of fast-spectrum and moderated space reactors separated into three areas—technical risk, performance, and safety/safeguards. A mix of quantitative and qualitative arguments is presented, and some conclusions generally can be made regarding neutron spectrum and space power application. In most cases, a fast-spectrum system appears to be the better alternative (mostly because of simplicity and higher potential operating temperatures); however, in some cases, such as a low-power (<100-kWt) surface reactor, a moderated spectrum could provide a better approach. In all cases, the determination of which spectrum is preferred is a strong function of the metrics provided by the "customer"— i.e., if a certain level of performance is required, it could provide a different solution than if a certain level of safeguards is required (which in some cases could produce a null solution). The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

On fundamental quality of fission chain reaction to oppose rapid runaways of nuclear reactors

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.

2017-01-01

It has been shown that the in-hour equation characterizes the barriers and resistibility of fission chain reaction (FCR) against rapid runaways in nuclear reactors. Traditionally, nuclear reactors are characterized by the presence of barriers based on delayed and prompt neutrons. A new barrier based on the reflector neutrons that can occur when the fast reactor core is surrounded by a weakly absorbing neutron reflector with heavy atomic weight was proposed. It has been shown that the safety of this fast reactor is substantially improved, and considerable elongation of prompt neutron lifetime "devalues" the role of delayed neutron fraction as the maximum permissible reactivity for the reactor safety.

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

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

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

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Preliminary Options Assessment of Versatile Irradiation Test Reactor

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

Sen, Ramazan Sonat

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 xmore » 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.« less

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

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

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

Slow clean-up for fast reactor

NASA Astrophysics Data System (ADS)

Banks, Michael

2008-05-01

The year 2300 is so distant that one may be forgiven for thinking of it only in terms of science fiction. But this is the year that workers at the Dounreay power station in Northern Scotland - the UK's only centre for research into "fast" nuclear reactors - term as the "end point" by which time the site will be completely clear of radioactive material. More than 180 facilities - including the iconic dome that housed the Dounreay Fast Reactor (DFR) - were built at at the site since it opened in 1959, with almost 50 having been used to handle radioactive material.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

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

DTIC Science & Technology

2010-07-09

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

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

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

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

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

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

NASA Astrophysics Data System (ADS)

Blandinskiy, V. Yu.

2014-12-01

This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

Prediction of the thermophysical properties of molten salt fast reactor fuel from first-principles

NASA Astrophysics Data System (ADS)

Gheribi, A. E.; Corradini, D.; Dewan, L.; Chartrand, P.; Simon, C.; Madden, P. A.; Salanne, M.

2014-05-01

Molten fluorides are known to show favourable thermophysical properties which make them good candidate coolants for nuclear fission reactors. Here we investigate the special case of mixtures of lithium fluoride and thorium fluoride, which act both as coolant and as fuel in the molten salt fast reactor concept. By using ab initio parameterised polarisable force fields, we show that it is possible to calculate the whole set of properties (density, thermal expansion, heat capacity, viscosity and thermal conductivity) which are necessary for assessing the heat transfer performance of the melt over the whole range of compositions and temperatures. We then deduce from our calculations several figures of merit which are important in helping the optimisation of the design of molten salt fast reactors.

Secure FAST: Security Enhancement in the NATO Time Sensitive Targeting Tool

DTIC Science & Technology

2010-11-01

designed to aid in the tracking and prosecuting of Time Sensitive Targets. The FAST tool provides user level authentication and authorisation in terms...level authentication and authorisation in terms of security. It uses operating system level security but does not provide application level security for...and collaboration tool, designed to aid in the tracking and prosecuting of Time Sensitive Targets. The FAST tool provides user level authentication and

78 FR 69139 - Physical Security-Design Certification and Operating Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-18

... scheduled to close on October 30, 2013. The Nuclear Energy Institute (NEI) submitted a letter on October 9... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan--draft section...

Space radiation studies at the White Sands Missile Range Fast Burst Reactor

NASA Technical Reports Server (NTRS)

Delapaz, A.

1972-01-01

The operation of the White Sands Missile Range Fast Burst Reactor is discussed. Space radiation studies in radiobiology, dosimetry, and transient radiation effects on electronic systems and components are described. Proposed modifications to increase the capability of the facility are discussed.

FAST CHOPPER BUILDING, TRA665. DETAIL OF STEEL DOOR ENTRY TO ...

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

FAST CHOPPER BUILDING, TRA-665. DETAIL OF STEEL DOOR ENTRY TO LOWER LEVEL. CAMERA FACING NORTH. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

DOPPLER CALCULATIONS FOR LARGE FAST CERAMIC REACTORS--EFFECTS OF IMPROVED METHODS AND RECENT CROSS SECTION INFORMATION

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

Greebler, P.; Goldman, E.

1962-12-19

Doppler calculations for large fast ceramic reactors (FCR), using recent cross section information and improved methods, are described. Cross sections of U/sup 238/, Pu/sup 239/, and Pu/sup 210/ with fuel temperature variations needed for perturbation calculations of Doppler reactivity changes are tabulated as a function of potential scattering cross section per absorber isotope at energies below 400 kev. These may be used in Doppler calculations for anv fast reactor. Results of Doppler calculations on a large fast ceramic reactor are given to show the effects of the improved calculation methods and of recent cross secrion data on the calculated Dopplermore » coefficient. The updated methods and cross sections used yield a somewhat harder spectrum and accordingly a somewhat smaller Doppler coefficient for a given FCR core size and composition than calculated in earlier work, but they support the essential conclusion derived earlier that the Doppler effect provides an important safety advantage in a large FCR. 28 references. (auth)« less

Impact of Including Higher Actinides in Fast Reactor Transmutation Analyses

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

B. Forget; M. Asgari; R. Ferrer

2007-09-01

Previous fast reactor transmutation studies generally disregarded higher mass minor actinides beyond Cm-246 due to various considerations including deficiencies in nuclear cross-section data. Although omission of these higher mass actinides does not significantly impact the neutronic calculations and fuel cycle performance parameters follow-on neutron dose calculations related to fuel recycling, transportation and handling are significantly impacted. This report shows that including the minor actinides in the equilibrium fast reactor calculations will increase the predicted neutron emission by about 30%. In addition a sensitivity study was initiated by comparing the impact of different cross-section evaluation file for representing these minor actinides.

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

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

Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.

1987-11-01

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alphamore » concentration significantly influence the ignition and steady-state burn capability. 23 refs., 9 figs., 4 tabs.« less

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

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

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

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

KINETICS OF TREAT USED AS A TEST REACTOR

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

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

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

Chen, J.; Alpan, F. A.; Fischer, G.A.

2011-07-01

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locationsmore » and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)« less

SAFETY AND SECURITY BUILDING, TRA614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. ...

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

SAFETY AND SECURITY BUILDING, TRA-614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. BLAW-KNOX 3150-814-2, 3/1950. INL INDEX NO. 531-0614-00-098-100703, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SAFETY AND SECURITY BUILDING, TRA614. SIMPLIFIED FLOOR LAYOUT AND WEST ...

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

SAFETY AND SECURITY BUILDING, TRA-614. SIMPLIFIED FLOOR LAYOUT AND WEST ELEVATION. BLAW-KNOX 3150-14-1, 1/1950. INL INDEX NO. 531-0614-00-098-100024, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

DOEpatents

Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

1962-06-26

A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

CONTROL SYSTEM FOR NEUTRONIC REACTORS

DOEpatents

Crever, F.E.

1962-05-01

BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

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

2009-07-01

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

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

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

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Bio-oil production from palm fronds by fast pyrolysis process in fluidized bed reactor

NASA Astrophysics Data System (ADS)

Rinaldi, Nino; Simanungkalit, Sabar P.; Kiky Corneliasari, S.

2017-01-01

Fast pyrolysis process of palm fronds has been conducted in the fluidized bed reactor to yield bio-oil product (pyrolysis oil). The process employed sea sand as the heat transfer medium. The objective of this study is to design of the fluidized bed rector, to conduct fast pyrolysis process to product bio-oil from palm fronds, and to characterize the feed and bio-oil product. The fast pyrolysis process was conducted continuously with the feeding rate around 500 g/hr. It was found that the biomass conversion is about 35.5% to yield bio-oil, however this conversion is still minor. It is suggested due to the heating system inside the reactor was not enough to decompose the palm fronds as a feedstock. Moreover, the acids compounds ware mostly observed on the bio-oil product.

3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

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

Ku, J.H.; Choung, W.M.; You, G.S.

The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluationmore » plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.« less

Fast-quench reactor for hydrogen and elemental carbon production from natural gas and other hydrocarbons

DOEpatents

Detering, Brent A.; Kong, Peter C.

2006-08-29

A fast-quench reactor for production of diatomic hydrogen and unsaturated carbons is provided. During the fast quench in the downstream diverging section of the nozzle, such as in a free expansion chamber, the unsaturated hydrocarbons are further decomposed by reheating the reactor gases. More diatomic hydrogen is produced, along with elemental carbon. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The product is a substantially clean-burning hydrogen fuel that leaves no greenhouse gas emissions, and elemental carbon that may be used in powder form as a commodity for several processes.

Fast quench reactor method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

1999-08-10

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

Fast quench reactor method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

1999-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

CHARACTERISTICS OF A FAST RISE TIME POWER SUPPLY FOR A PULSED PLASMA REACTOR FOR CHEMICAL VAPOR DESTRUCTION

EPA Science Inventory

Rotating spark gap devices for switching high-voltage direct current (dc) into a corona plasma reactor can achieve pulse rise times in the range of tens of nanoseconds. The fast rise times lead to vigorous plasma generation without sparking at instantaneous applied voltages highe...

The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

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

Smith, R.R.

1986-01-01

This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

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

Furuta, H.; Imura, A.; Furuta, Y.

Recently, technique of Gadolinium loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and 'nuclear Gain (GA)' for IAEA safeguards. For the practical use, R and D of the 1 ton class compact detector, which is measurable above ground, is necessary. Especially, it is important to reduce much amount of fast neutron background induced by cosmic muons with data analysis for the measurement above ground. We developed a prototype of the Gd-LS detector with 200 L of the target volume, which has Pulse Shape Discrimination (PSD) ability for the fast neutronmore » reduction with data analysis. Usually, it is well known that it is difficult to keep high fast neutron reduction power of PSD with the large volume size such as the neutrino reactor monitor. We evaluated the PSD ability of our prototype with real fast neutrons induced by the muons in our laboratory above ground, and we could confirm to keep the high fast neutron reduction power with even our large detector size. (authors)« less

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

A roadmap for nuclear energy technology

NASA Astrophysics Data System (ADS)

Sofu, Tanju

2018-01-01

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

Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

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

Stoller, R.E.

1987-12-01

The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and alteredmore » mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.« less

NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

DOEpatents

Mills, F.T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

Neutronic Reactor Design to Reduce Neutron Loss

DOEpatents

Miles, F. T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

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

NASA Astrophysics Data System (ADS)

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

2015-12-01

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

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

Fuel Cycle System Analysis Handbook

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

Steven J. Piet; Brent W. Dixon; Dirk Gombert

2009-06-01

This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some ofmore » the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty diagrams, which show at a glance combined uncertainty information, section 9.2 has a new set of simpler graphs that show the impact on fuel cycle costs for once through, 1-tier, and 2-tier scenarios as a function of key input parameters.« less

SAFETY AND SECURITY BUILDING, TRA614. FLOOR, ROOF, AND FOUNDATION PLANS. ...

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

SAFETY AND SECURITY BUILDING, TRA-614. FLOOR, ROOF, AND FOUNDATION PLANS. ROOM FUNCTIONS. DOOR AND ROOM FINISH SCHEDULE. BLAW-KNOX 3150-814-1, 3/1950. INL INDEX NO. 531-0614-00-098-100702, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The slow and fast pyrolysis of cherry seed.

PubMed

Duman, Gozde; Okutucu, Cagdas; Ucar, Suat; Stahl, Ralph; Yanik, Jale

2011-01-01

The slow and fast pyrolysis of cherry seeds (CWS) and cherry seeds shells (CSS) was studied in fixed-bed and fluidized bed reactors at different pyrolysis temperatures. The effects of reactor type and temperature on the yields and composition of products were investigated. In the case of fast pyrolysis, the maximum bio-oil yield was found to be about 44 wt% at pyrolysis temperature of 500 °C for both CWS and CSS, whereas the bio yields were of 21 and 15 wt% obtained at 500 °C from slow pyrolysis of CWS and CSS, respectively. Both temperature and reactor type affected the composition of bio-oils. The results showed that bio-oils obtained from slow pyrolysis of CWS and CSS can be used as a fuel for combustion systems in industry and the bio-oil produced from fast pyrolysis can be evaluated as a chemical feedstock. Copyright © 2010 Elsevier Ltd. All rights reserved.

Shielded fluid stream injector for particle bed reactor

DOEpatents

Notestein, John E.

1993-01-01

A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

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

Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

2007-11-01

As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinidesmore » above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.« less

Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

NASA Astrophysics Data System (ADS)

Dutta, N. G.

2012-11-01

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

Nuclear reactor downcomer flow deflector

DOEpatents

Gilmore, Charles B [Greensburg, PA; Altman, David A [Pittsburgh, PA; Singleton, Norman R [Murrysville, PA

2011-02-15

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

Radial blanket assembly orificing arrangement

DOEpatents

Patterson, J.F.

1975-07-01

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

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

Hess, J. D.; Gauld, I. C.; Gulliford, J.

2017-01-01

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Projectmore » (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.« less

Catalytic fast pyrolysis of white oak wood in-situ using a bubbling fluidized bed reactor

USDA-ARS?s Scientific Manuscript database

Catalytic fast pyrolysis was performed on white oak wood using two zeolite-type catalysts as bed material in a bubbling fluidized bed reactor. The two catalysts chosen, based on a previous screening study, were Ca2+ exchanged Y54 (Ca-Y54) and a proprietary ß-zeolite type catalyst (catalyst M) both ...

FAST CHOPPER BUILDING, TRA665. DETAIL SHOWS UPPER AND LOWER LEVEL ...

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

FAST CHOPPER BUILDING, TRA-665. DETAIL SHOWS UPPER AND LOWER LEVEL WALLS OF DIFFERING MATERIALS. NOTE DOORWAY TO MTR TO RIGHT OF CHOPPER BUILDING'S CLIPPED CORNER. CAMERA FACING WEST. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

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

Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

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

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

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

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

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

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

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

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.

FAST NEUTRON REACTOR

DOEpatents

Soodak, H.; Wigner, E.P.

1961-07-25

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

76 FR 72007 - ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-21

...; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor... activities in nuclear power reactors against radiological sabotage,'' paragraph (b)(1) states, ``The licensee...

75 FR 29588 - Office of New Reactors: Proposed NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-26

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0184] Office of New Reactors: Proposed NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision 0 on Cyber Security Plan AGENCY: Nuclear Regulatory... Plants,'' on a proposed Standard Review Plan (SRP) Section 13.6.6 on ``Cyber Security Plan'' (Agencywide...

Hydrogen and elemental carbon production from natural gas and other hydrocarbons

DOEpatents

Detering, Brent A.; Kong, Peter C.

2002-01-01

Diatomic hydrogen and unsaturated hydrocarbons are produced as reactor gases in a fast quench reactor. During the fast quench, the unsaturated hydrocarbons are further decomposed by reheating the reactor gases. More diatomic hydrogen is produced, along with elemental carbon. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The product is a substantially clean-burning hydrogen fuel that leaves no greenhouse gas emissions, and elemental carbon that may be used in powder form as a commodity for several processes.

The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

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

Anderson, Robert Stephen

Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologiesmore » may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.« less

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

Packed rod neutron shield for fast nuclear reactors

DOEpatents

Eck, John E.; Kasberg, Alvin H.

1978-01-01

A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

FAST CHOPPER DETECTOR HOUSE, TRA665. SECOND FLOOR ADDITION: PLAN, SECTIONS ...

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

FAST CHOPPER DETECTOR HOUSE, TRA-665. SECOND FLOOR ADDITION: PLAN, SECTIONS AND DETAILS AS ADDED TO THE EXISTING CHOPPER HOUSE IN 1962. F.C. TORKELSON 842-MTR-665-S-3, 4/1962. INL INDEX NO. 531-0665-60-851-150997, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

PubMed

Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n∕cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

NASA Astrophysics Data System (ADS)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

A summary of sodium-cooled fast reactor development

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

Aoto, Kazumi; Dufour, Philippe; Hongyi, Yang

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicatedmore » to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.« less

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

NASA Astrophysics Data System (ADS)

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

2012-08-01

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

Lessons Learned about Liquid Metal Reactors from FFTF Experience

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

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test the robustness of the conclusions presented in the MIT Fuel Cycle Study. These conclusions are found to still hold, even when considering alternative technologies and different sets of simulation assumptions. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. Optimization metrics of interest for different stakeholders in the fuel cycle (economics, fuel resource utilization, high level waste, transuranics/proliferation management, and environmental impact) are utilized for two different optimization techniques: a linear one and a stochastic one. Stakeholder elicitation provided sets of relative weights for the identified metrics appropriate to each stakeholder group, which were then successfully used to arrive at optimum fuel cycle configurations for recycling technologies. The stochastic optimization tool, based on a genetic algorithm, was used to identify non-inferior solutions according to Pareto's dominance approach to optimization. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs mit.edu)

Cross-Section Measurements in the Fast Neutron Energy Range

NASA Astrophysics Data System (ADS)

Plompen, Arjan

2006-04-01

Generation IV focuses research for advanced nuclear reactors on six concepts. Three of these concepts, the lead, gas and sodium fast reactors (LFR, GFR and SFR) have fast neutron spectra, whereas a fourth, the super-critical water reactor (SCWR), can be configured to have a fast spectrum. Such fast neutron spectra are essential to meet the sustainability objective of GenIV. Nuclear data requirements for GenIV concepts will therefore emphasize the energy region from about 1 keV to 10 MeV. Here, the potential is illustrated of the GELINA neutron time-of-flight facility and the Van de Graaff laboratory at IRMM to measure the relevant nuclear data in this energy range: the total, capture, fission and inelastic-scattering cross sections. In particular, measurement results will be shown for lead and bismuth inelastic scattering for which the need was recently expressed in a quantitative way by Aliberti et al. for Accelerator Driven Systems. Even without completion of the quantitative assessment of the data needs for GenIV concepts at ANL it is clear that this particular effort is of relevance to LFR system studies.

Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

NASA Astrophysics Data System (ADS)

Bays, Samuel Eugene

2008-10-01

In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this fast fissile quality and also the fact that Pu-238 is transmuted from Np-237 and Am-241, these MAs are regarded as fertile material in the SFR design proposed by this dissertation. This dissertation demonstrates a SFR design which is dedicated to plutonium breeding by targeting Am-241 transmutation. This SFR design uses a moderated axial transmutation target that functions primarily as a pseudo-blanket fuel, which is reprocessed with the active driver fuel in an integrated recycling strategy. This work demonstrates the cost and feasibility advantages of plutonium breeding via MA transmutation by adopting reactor, reprocessing and fuel technologies previously demonstrated for traditional breeder reactors. The fuel cycle proposed seeks to find a harmony between the waste management advantages of transuranic burning SFRs and the resource sustainability of traditional plutonium breeder SFRs. As a result, the enhanced plutonium conversion from MAs decreases the burner SFR's fuel costs, by extracting more fissile value from the initial TRU purchased through SNF reprocessing.

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

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

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

FAST CHOPPER DETECTOR HOUSE, TRA665. FIRST FLOOR, PLAN AND SECTION, ...

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

FAST CHOPPER DETECTOR HOUSE, TRA-665. FIRST FLOOR, PLAN AND SECTION, AS PROPOSED FOR MODIFICATION IN 1962. CONCRETE WALLS THREE FEET THICK. EXISTING WINDOWS IN MTR AND DETECTOR HOUSE WALLS WERE TO BE FILLED IN WITH HIGH-DENSITY BRICK. NOTE 20-METER MARK, WHERE THE FAST CHOPPER DETECTOR HAD BEEN LOCATED. F.C. TORKELSON 842-MTR-665-S-2, 4/1962. INL INDEX NO. 531-0665-60-851-150996, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

DOEpatents

Brehm, Jr., William F.; Colburn, Richard P.

1982-01-01

An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

Status of liquid metal fast breeder reactor fuel development in Japan

NASA Astrophysics Data System (ADS)

Katsuragawa, M.; Kashihara, H.; Akebi, M.

1993-09-01

The mixed-oxide fuel technology for a liquid metal fast breeder reactor (LMFBR) in Japan is progressing toward commercial deployment of LMFBR. Based on accumulated experience in Joyo and Monju fuel development, efforts for large scale LMFBR fuel development are devoted to improved irradiation performance, reliability and economy. This paper summarizes accomplishments, current activities and future plans for LMFBR fuel development in Japan.

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)

FAST CHOPPER BUILDING, TRA665, INTERIOR. LOWER (DETECTOR) LEVEL. NOTE BRICKEDIN ...

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

FAST CHOPPER BUILDING, TRA-665, INTERIOR. LOWER (DETECTOR) LEVEL. NOTE BRICKED-IN WINDOW ON MTR SIDE. USED FOR STORAGE OF LEAD BRICKS AFTER EXPERIMENTAL NEUTRON INSTRUMENTS WERE REMOVED. SIGN SAYS "IN-PROCESS LEAD SOURCE STORAGE." INL NEGATIVE NO. HD-42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

New Technological Platform for the National Nuclear Energy Strategy Development

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Rachkov, V. I.

2017-12-01

The paper considers the need to update the development strategy of Russia's nuclear power industry and various approaches to the large-scale nuclear power development. Problems of making decisions on fast neutron reactors and closed nuclear fuel cycle (NFC) arrangement are discussed. The current state of the development of fast neutron reactors and closed NFC technologies in Russia is considered and major problems are highlighted.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

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

Koch, Len; Agnew, Harold

Dr. Harold Agnew, retired director of Los Alamos National Laboratory and one of 49 people present on December 2, 1942 when the world’s first man-made controlled nuclear chain reaction was achieved with the CP-1 reactor, and Dr. Len Koch, one of Argonne’s earliest staff members and a designer of EBR-I, the first liquid metal-cooled fast reactor, spoke about their early work during “The Dawn of the Nuclear Age”, a Director’s Special Symposium held as one of the events to commemorate the 70th anniversary year of CP-1 achieving criticality. The symposium was moderated by Dr. Charles Till, a retired Argonne associatemore » laboratory director who led Argonne’s nuclear engineering programs throughout the 1980’s and ‘90’s. Dr. Agnew painted a vivid picture of the challenges and rewards of working in Enrico Fermi’s group under strict security conditions and the complete faith all in the group had in Fermi’s analyses. He stated that no one ever doubted that CP-1 would achieve criticality, and when the moment came, those present acknowledged the accomplishment with little more than a subdued toast of chianti from a bottle provided by reactor physicist Eugene Wigner. This experimental work on nuclear reactors was continued in the Chicago area and led first by Fermi and then Walter Zinn, another member of Fermi’s CP-1 group, resulting in the formal establishment of Argonne National Laboratory on July 1, 1946. Dr. Koch described how much he enjoyed working at Argonne through the 1950’s and ‘60’s and contributing to many of the research “firsts” that Argonne achieved in the nuclear energy field and led to the foundation of the commercial nuclear power generation industry. His reminiscences about all that was achieved with EBR-I and how that work then led into Argonne’s design, building, and operation of EBR-II as a full demonstration of a fast reactor power plant brought Argonne’s nuclear energy legacy to life for everyone in the auditorium. These engaging presentations prompted a number of questions from the audience, which showed its appreciation for both speakers by ending the symposium with a standing ovation.« less

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

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

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

1978-06-01

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

Continuous production of tritium in an isotope-production reactor with a separate circulation system

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

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.

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

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

1998-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

Fast Breeder Reactors in Sweden: Vision and Reality.

PubMed

Fjaestad, Maja

2015-01-01

The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change.

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

Unneberg, L.

The main features of the 16 core grids (top guides) designed by ABB ATOM AB are briefly described and the evolution of the design is discussed. One important characteristic of the first nine grids is the existence of bolts securing guide bars to the core grid plates. These bolts are made of precipitation hardened or solution annealed stainless steel. During operation, bolts in all none grids have cracked. The failure analyses indicate that intergranular stress corrosion cracking (IGSCC), possibly accelerated by crevice conditions and/or irradiation, was the cause of failure. Fast neutron fluences approaching or exceeding the levels considered asmore » critical for irradiation assisted stress corrosion cracking (IASCC) will be reached in a few cases only. Temporary measures were taken immediately after the discovery of the cracking. For five of the nine reactors affected, it was decided to replace the complete grids. Two of these replacements have been successfully carried out to date. IASCC as a potential future problem is discussed and it is pointed out that, during their life times, the ABB ATOM core grids will be exposed to sufficiently high fast neutron fluences to cause some concern.« less

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering Laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems. (author)« less

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

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

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

Steam generator for liquid metal fast breeder reactor

DOEpatents

Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

1985-01-01

Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

NASA Astrophysics Data System (ADS)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-30

... the Standard Review Plan (SRP), concerning the physical security reviews of design certification... NRC staff with the physical security review of applications for design certifications, incorporate... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Proposed Revision to Physical Security--Standard...

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.

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

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

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

Fanning, T. H.

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

NEUTRONIC REACTOR

DOEpatents

Wade, E.J.

1958-09-16

This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

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

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

An Ultrasonic Scanning Technique for In-Situ `Bowing' Measurement of Prototype Fast Breeder Reactor Fuel Sub-Assembly

NASA Astrophysics Data System (ADS)

Swaminathan, K.; Asokane, C.; Sylvia, J. I.; Kalyanasundaram, P.; Swaminathan, P.

2012-02-01

An ultrasonic under-sodium scanner has been developed for deployment in Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India. Its purpose is to scan the above-core plenum for detection, if any, of displacement of sub-assemblies. During its burn-up in the reactor, the head of a Fuel Sub-Assembly (FSA) may undergo a lateral shift from its original position (called `bowing') due to the fast neutron induced damage on its structural material. A simple scanning technique has been developed for measuring the extent of bowing in-situ. This paper describes a PC-controlled mock-up of the scanner used to implement the scanning technique and the results obtained of scanning a mock-up FSA head under water. The details of the liquid-sodium proof transducer developed for use in the PFBR scanner and its performance are also discussed.

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

Bustraan, M.; Coehoorn, J.; Veenema, J.J.

This report is a collection of separate contributions on some aspects of the work done on STEK up to February 1970. A description is given of STEK together with the philosophy of its design, i.e. integral measurements of fission product cross sections by a sample oscillator technique in fast reactor spectra. The influences fission products may have on fast breeder reactors are briefly demonstrated by an example. A description of the facility and of the sample oscillator and sample exchange mechanism is given. Some preliminary results of measurements of reactor parameters and the neutron spectrum in the first fast zonemore » in STEK are given. For the use of lead as material for the buffer an argumentation is given. The proposed program for the measurements of the integral fission product cross sections is outlined. The procurement of some, highly active, samples of actual fission products is briefly sketched. (auth)« less

Estimating the Temperature Experienced by Biomass Particles during Fast Pyrolysis Using Microscopic Analysis of Biochars

DOE PAGES

Thompson, Logan C.; Ciesielski, Peter N.; Jarvis, Mark W.; ...

2017-07-12

Here, biomass particles can experience variable thermal conditions during fast pyrolysis due to differences in their size and morphology, and from local temperature variations within a reactor. These differences lead to increased heterogeneity of the chemical products obtained in the pyrolysis vapors and bio-oil. Here we present a simple, high-throughput method to investigate the thermal history experienced by large ensembles of particles during fast pyrolysis by imaging and quantitative image analysis. We present a correlation between the surface luminance (darkness) of the biochar particle and the highest temperature that it experienced during pyrolysis. Next, we apply this correlation to large,more » heterogeneous ensembles of char particles produced in a laminar entrained flow reactor (LEFR). The results are used to interpret the actual temperature distributions delivered by the reactor over a range of operating conditions.« less

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

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

DOE PAGES

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

2015-05-21

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

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.

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

75 FR 7641 - Self-Regulatory Organizations; The Depository Trust Company; Notice of Filing of Proposed Rule...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-22

... Fast Automated Securities Transfer Program February 5, 2010. Pursuant to Section 19(b)(1) of the... American Depositary Receipts (``ADRs'') that are a part of DTC's Fast Automated Securities Transfer Program (``FAST''). II. Self-Regulatory Organization's Statement of the Purpose of, and Statutory Basis for, the...

A Security Analysis on Kempf-Koodli's Security Scheme for Fast Mobile IPv6

NASA Astrophysics Data System (ADS)

You, Ilsun; Sakurai, Kouichi; Hori, Yoshiaki

Recently, the security scheme, proposed by Kempf and Koodli, has been adopted as a security standard for Fast handover for Mobile IPv6. But, it does not prevent denial of service attacks while resulting in high computation cost. More importantly, we find that it is still vulnerable to redirection attacks because it fails to secure the Unsolicited Neighbor Advertisement messages. In this paper, Kempf-Koodli's scheme is formally analyzed through BAN-logic and its weaknesses are demonstrated.

Applications in Nuclear Energy Security

NASA Astrophysics Data System (ADS)

Sheffield, Richard

2009-05-01

A key roadblock to development of additional nuclear power capacity is a concern over management of nuclear waste. Nuclear waste is predominantly comprised of used fuel discharged from operating nuclear reactors. The roughly 100 operating US reactors currently produce about 20% of the US electricity and will create about 87,000 tons of such discharged or ``spent'' fuel over the course of their lifetimes. The long-term radioactivity of the spent fuel drives the need for deep geologic storage that remains stable for millions of years. Nearly all issues related to risks to future generations arising from long-term disposal of such spent nuclear fuel is attributable to approximately the 1% made up primarily of minor actinides. If we can reduce or eliminate this 1% of the spent fuel, then within a few hundred years the toxic nature of the spent fuel drops below that of the natural uranium ore that was originally mined for nuclear fuel. The minor actinides can be efficiently eliminated through nuclear transmutation using as a driver fast-neutrons produced by a spallation process initiated with a high-energy proton beam. This presentation will cover the system design considerations and issues of an accelerator driven transmutation system.

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

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

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

Introduction to Reactor Statics Modules, RS-1. Nuclear Engineering Computer Modules.

ERIC Educational Resources Information Center

Edlund, Milton C.

The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burn-up for both slow neutron and fast neutron fission reactors. The diffusion…

Waste tyre pyrolysis: modelling of a moving bed reactor.

PubMed

Aylón, E; Fernández-Colino, A; Murillo, R; Grasa, G; Navarro, M V; García, T; Mastral, A M

2010-12-01

This paper describes the development of a new model for waste tyre pyrolysis in a moving bed reactor. This model comprises three different sub-models: a kinetic sub-model that predicts solid conversion in terms of reaction time and temperature, a heat transfer sub-model that calculates the temperature profile inside the particle and the energy flux from the surroundings to the tyre particles and, finally, a hydrodynamic model that predicts the solid flow pattern inside the reactor. These three sub-models have been integrated in order to develop a comprehensive reactor model. Experimental results were obtained in a continuous moving bed reactor and used to validate model predictions, with good approximation achieved between the experimental and simulated results. In addition, a parametric study of the model was carried out, which showed that tyre particle heating is clearly faster than average particle residence time inside the reactor. Therefore, this fast particle heating together with fast reaction kinetics enables total solid conversion to be achieved in this system in accordance with the predictive model. Copyright © 2010 Elsevier Ltd. All rights reserved.

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

Mass tracking and material accounting in the integral fast reactor (IFR)

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

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstratedmore » in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.« less

FAST NEUTRONIC REACTOR

DOEpatents

Snell, A.H.

1957-12-01

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

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

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

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

FAST CHOPPER BUILDING, TRA665. CAMERA FACING NORTH. NOTE BRICKEDIN WINDOW ...

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

FAST CHOPPER BUILDING, TRA-665. CAMERA FACING NORTH. NOTE BRICKED-IN WINDOW ON RIGHT SIDE (BELOW PAINTED NUMERALS "665"). SLIDING METAL DOOR ON COVERED RAIL AT UPPER LEVEL. SHELTERED ENTRANCE TO STEEL SHIELDING DOOR. DOOR INTO MTR SERVICE BUILDING, TRA-635, STANDS OPEN. MTR BEHIND CHOPPER BUILDING. INL NEGATIVE NO. HD42-1. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Use of freeze-casting in advanced burner reactor fuel design

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

Lang, A. L.; Yablinsky, C. A.; Allen, T. R.

2012-07-01

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by thatmore » fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)« less

Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

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

Osipenko, M.; Ripani, M.; Ricco, G.

2015-07-01

A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based onmore » conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)« less

An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions.

PubMed

Ruan, Guihua; Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui; Du, Fuyou

2016-04-22

A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of Nα-benzoyl-l-arginine ethyl ester to Nα-benzoyl-l-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. Copyright © 2016 Elsevier Inc. All rights reserved.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

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

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

Nuclear reactor shield including magnesium oxide

DOEpatents

Rouse, Carl A.; Simnad, Massoud T.

1981-01-01

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Response of the first wetted wall of an IFE reactor chamber to the energy release from a direct-drive DT capsule

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

Medin, Stanislav A.; Basko, Mikhail M.; Orlov, Yurii N.

2012-07-11

Radiation hydrodynamics 1D simulations were performed with two concurrent codes, DEIRA and RAMPHY. The DEIRA code was used for DT capsule implosion and burn, and the RAMPHY code was used for computation of X-ray and fast ions deposition in the first wall liquid film of the reactor chamber. The simulations were run for 740 MJ direct drive DT capsule and Pb thin liquid wall reactor chamber of 10 m diameter. Temporal profiles for DT capsule leaking power of X-rays, neutrons and fast {sup 4}He ions were obtained and spatial profiles of the liquid film flow parameter were computed and analyzed.

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)

Nuclear instrumentation in VENUS-F

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Apparatus for detecting leakage of liquid sodium

DOEpatents

Himeno, Yoshiaki

1978-01-01

An apparatus for detecting the leakage of liquid sodium includes a cable-like sensor adapted to be secured to a wall of piping or other equipment having sodium on the opposite side of the wall, and the sensor includes a core wire electrically connected to the wall through a leak current detector and a power source. An accidental leakage of the liquid sodium causes the corrosion of a metallic layer and an insulative layer of the sensor by products resulted from a reaction of sodium with water or oxygen in the atmospheric air so as to decrease the resistance between the core wire and the wall. Thus, the leakage is detected as an increase in the leaking electrical current. The apparatus is especially adapted for use in detecting the leakage of liquid sodium from sodium-conveying pipes or equipment in a fast breeder reactor.

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.

Small-scale nuclear reactors for remote military operations: opportunities and challenges

DTIC Science & Technology

2015-08-25

study – Report was published in March 2011  CNA study identified challenges to deploy small modular reactors (SMRs) at a base – Identified First-of...forward operating bases. The availability of deployable, cost-effective, regulated, and secure small modular reactors with a modest output electrical...defense committees on the challenges, operational requirements, constraints, cost, and life cycle analysis for a small modular reactor of less than 10

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

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

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

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

Bucknor, M.; Farmer, M.; Grabaskas, D.

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination ofmore » gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.« less

Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

NASA Astrophysics Data System (ADS)

Krasikov, E.; Nikolaenko, V.

2017-01-01

Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

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

DOE PAGES

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

2016-01-01

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

Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

PubMed

Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

2017-10-24

Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (2) Critical experiment of lithium-6 used in LEM and LIM

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

Tsunoda, Hirokazu; Sato, Osamu; Okajima, Shigeaki

2002-07-01

In order to achieve fully automated reactor operation of RAPID-L reactor, innovative reactivity control systems LEM, LIM, and LRM are equipped with lithium-6 as a liquid poison. Because lithium-6 has not been used as a neutron absorbing material of conventional fast reactors, measurements of the reactivity worth of Lithium-6 were performed at the Fast Critical Assembly (FCA) of Japan Atomic Energy Research Institute (JAERI). The FCA core was composed of highly enriched uranium and stainless steel samples so as to simulate the core spectrum of RAPID-L. The samples of 95% enriched lithium-6 were inserted into the core parallel to themore » core axis for the measurement of the reactivity worth at each position. It was found that the measured reactivity worth in the core region well agreed with calculated value by the method for the core designs of RAPID-L. Bias factors for the core design method were obtained by comparing between experimental and calculated results. The factors were used to determine the number of LEM and LIM equipped in the core to achieve fully automated operation of RAPID-L. (authors)« less

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

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

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-06-22

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluencemore » monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.« less

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

DOE PAGES

Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; ...

2015-06-18

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. Themore » particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.« less

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

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

S. Khericha

2010-12-01

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn thesemore » actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.« less

SAM Theory Manual

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

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Ochoa, R.; Martínez, J. S.; Díez, C. J.; García-Herranz, N.; Cabellos, O.

2014-04-01

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

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

Jiao, Zhijie

The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

Locked-wrap fuel rod

DOEpatents

Kaplan, Samuel; Chertock, Alan J.; Punches, James R.

1977-01-01

A method for spacing fast reactor fuel rods using a wire wrapper improved by orienting the wire-wrapped fuel rods in a unique manner which introduces desirable performance characteristics not attainable by previous wire-wrapped designs. Use of this method in a liquid metal fast breeder reactor results in: (a) improved mechanical performance, (b) improved rod-to-rod contact, (c) reduced steel volume, and (d) improved thermal-hydraulic performance. The method produces a "locked wrap" design which tends to lock the rods together at each of the wire cluster locations.

Liquid metal fast breeder reactors, 1972--1973

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

None

1974-01-01

Reference to 1467 publications on liquid sodium fast breeder reactors cited in Nuclear Science Abstracts Volume 26 (1972) through Volume 27 (1973 through June) are contained in this citation to provide information on the contents of the document. References are arranged in order by the original NSA abstract number which approximately places them in chronological order. Sequence numbers appear beside each reference, and the personal author index refers to these sequence numbers. The subject index refers to the original abstract numbers. (auth)

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

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

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

Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

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

Malathi, N.; Sahoo, P., E-mail: sahoop@igcar.gov.in; Ananthanarayanan, R.

2015-02-15

An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision,more » sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.« less

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

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

Grabaskas, D.; Brunett, A.; Passerini, S.

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. Themore » mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.« less

Alternate Tritium Production Methods Using A Liquid Lithium Target

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

Wilson, J.

For over 60 years, the Savannah River Site’s primary mission has been the production of tritium. From the beginning, the Savannah River National Laboratory (SRNL) has provided the technical foundation to ensure the successful execution of this critical defense mission. SRNL has developed most of the processes used in the tritium mission and provides the research and development necessary to supply this critical component. This project was executed by first developing reactor models that could be used as a neutron source. In parallel to this development calculations were carried out testing the feasibility of accelerator technologies that could also bemore » used for tritium production. Targets were designed with internal moderating material and optimized target was calculated to be capable of 3000 grams using a 1400 MWt sodium fast reactor, 850 grams using a 400 MWt sodium fast reactor, and 100 grams using a 62 MWt reactor, annually.« less

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

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

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

Realizing "2001: A Space Odyssey": Piloted Spherical Torus Nuclear Fusion Propulsion

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Dudzinski, Leonard A.; Borowski, Stanley K.; Juhasz, Albert J.

2005-01-01

A conceptual vehicle design enabling fast, piloted outer solar system travel was created predicated on a small aspect ratio spherical torus nuclear fusion reactor. The initial requirements were satisfied by the vehicle concept, which could deliver a 172 mt crew payload from Earth to Jupiter rendezvous in 118 days, with an initial mass in low Earth orbit of 1,690 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including artificial gravity payload, central truss, nuclear fusion reactor, power conversion, magnetic nozzle, fast wave plasma heating, tankage, fuel pellet injector, startup/re-start fission reactor and battery bank, refrigeration, reaction control, communications, mission design, and space operations. Detailed fusion reactor design included analysis of plasma characteristics, power balance/utilization, first wall, toroidal field coils, heat transfer, and neutron/x-ray radiation. Technical comparisons are made between the vehicle concept and the interplanetary spacecraft depicted in the motion picture 2001: A Space Odyssey.

Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

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

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

Nuclear fuel requirements for the American economy - A model

NASA Astrophysics Data System (ADS)

Curtis, Thomas Dexter

A model is provided to determine the amounts of various fuel streams required to supply energy from planned and projected nuclear plant operations, including new builds. Flexible, user-defined scenarios can be constructed with respect to energy requirements, choices of reactors and choices of fuels. The model includes interactive effects and extends through 2099. Outputs include energy provided by reactors, the number of reactors, and masses of natural Uranium and other fuels used. Energy demand, including electricity and hydrogen, is obtained from US DOE historical data and projections, along with other studies of potential hydrogen demand. An option to include other energy demand to nuclear power is included. Reactor types modeled include (thermal reactors) PWRs, BWRs and MHRs and (fast reactors) GFRs and SFRs. The MHRs (VHTRs), GFRs and SFRs are similar to those described in the 2002 DOE "Roadmap for Generation IV Nuclear Energy Systems." Fuel source choices include natural Uranium, self-recycled spent fuel, Plutonium from breeder reactors and existing stockpiles of surplus HEU, military Plutonium, LWR spent fuel and depleted Uranium. Other reactors and fuel sources can be added to the model. Fidelity checks of the model's results indicate good agreement with historical Uranium use and number of reactors, and with DOE projections. The model supports conclusions that substantial use of natural Uranium will likely continue to the end of the 21st century, though legacy spent fuel and depleted uranium could easily supply all nuclear energy demand by shifting to predominant use of fast reactors.

75 FR 63867 - DTE Energy; Enrico Fermi Atomic Power Plant Unit 1, Exemption From Certain Security Requirements

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-18

... procedures, physical security plan, guard training and qualification plan, or cyber security plan for the... Power Plant Unit 1, Exemption From Certain Security Requirements 1.0 Background DTE Energy (DTE) is the... atmospheric pressure. In November 1972, the Power Reactor Development Company (PRDC), the licensee at that...

49 CFR 1572.201 - Transportation of hazardous materials via commercial motor vehicle from Canada or Mexico to and...

Code of Federal Regulations, 2013 CFR

2013-10-01

... subpart for purposes of this section: FAST means Free and Secure Trade program of the Bureau of Customs... 1572.5. (d) FAST card. A commercial motor vehicle driver who holds a current Free and Secure Trade (FAST) program card satisfies the requirements of this section. Commercial motor vehicle drivers who...

49 CFR 1572.201 - Transportation of hazardous materials via commercial motor vehicle from Canada or Mexico to and...

Code of Federal Regulations, 2011 CFR

2011-10-01

... subpart for purposes of this section: FAST means Free and Secure Trade program of the Bureau of Customs... 1572.5. (d) FAST card. A commercial motor vehicle driver who holds a current Free and Secure Trade (FAST) program card satisfies the requirements of this section. Commercial motor vehicle drivers who...

49 CFR 1572.201 - Transportation of hazardous materials via commercial motor vehicle from Canada or Mexico to and...

Code of Federal Regulations, 2012 CFR

2012-10-01

... subpart for purposes of this section: FAST means Free and Secure Trade program of the Bureau of Customs... 1572.5. (d) FAST card. A commercial motor vehicle driver who holds a current Free and Secure Trade (FAST) program card satisfies the requirements of this section. Commercial motor vehicle drivers who...

49 CFR 1572.201 - Transportation of hazardous materials via commercial motor vehicle from Canada or Mexico to and...

Code of Federal Regulations, 2014 CFR

2014-10-01

... subpart for purposes of this section: FAST means Free and Secure Trade program of the Bureau of Customs... 1572.5. (d) FAST card. A commercial motor vehicle driver who holds a current Free and Secure Trade (FAST) program card satisfies the requirements of this section. Commercial motor vehicle drivers who...

CP-1 70th Anniversary Symposium

ScienceCinema

Len Koch; Harold Agnew

2017-12-09

Dr. Harold Agnew, retired director of Los Alamos National Laboratory and one of 49 people present on December 2, 1942 when the world’s first man-made controlled nuclear chain reaction was achieved with the CP-1 reactor, and Dr. Len Koch, one of Argonne’s earliest staff members and a designer of EBR-I, the first liquid metal-cooled fast reactor, spoke about their early work during “The Dawn of the Nuclear Age”, a Director’s Special Symposium held as one of the events to commemorate the 70th anniversary year of CP-1 achieving criticality. The symposium was moderated by Dr. Charles Till, a retired Argonne associate laboratory director who led Argonne’s nuclear engineering programs throughout the 1980’s and ‘90’s. Dr. Agnew painted a vivid picture of the challenges and rewards of working in Enrico Fermi’s group under strict security conditions and the complete faith all in the group had in Fermi’s analyses. He stated that no one ever doubted that CP-1 would achieve criticality, and when the moment came, those present acknowledged the accomplishment with little more than a subdued toast of chianti from a bottle provided by reactor physicist Eugene Wigner. This experimental work on nuclear reactors was continued in the Chicago area and led first by Fermi and then Walter Zinn, another member of Fermi’s CP-1 group, resulting in the formal establishment of Argonne National Laboratory on July 1, 1946. Dr. Koch described how much he enjoyed working at Argonne through the 1950’s and ‘60’s and contributing to many of the research “firsts” that Argonne achieved in the nuclear energy field and led to the foundation of the commercial nuclear power generation industry. His reminiscences about all that was achieved with EBR-I and how that work then led into Argonne’s design, building, and operation of EBR-II as a full demonstration of a fast reactor power plant brought Argonne’s nuclear energy legacy to life for everyone in the auditorium. These engaging presentations prompted a number of questions from the audience, which showed its appreciation for both speakers by ending the symposium with a standing ovation.

Information basis for developing comprehensive waste management system-US-Japan joint nuclear energy action plan waste management working group phase I report.

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

Nutt, M.; Nuclear Engineering Division

2010-05-25

The activity of Phase I of the Waste Management Working Group under the United States - Japan Joint Nuclear Energy Action Plan started in 2007. The US-Japan JNEAP is a bilateral collaborative framework to support the global implementation of safe, secure, and sustainable, nuclear fuel cycles (referred to in this document as fuel cycles). The Waste Management Working Group was established by strong interest of both parties, which arise from the recognition that development and optimization of waste management and disposal system(s) are central issues of the present and future nuclear fuel cycles. This report summarizes the activity of themore » Waste Management Working Group that focused on consolidation of the existing technical basis between the U.S. and Japan and the joint development of a plan for future collaborative activities. Firstly, the political/regulatory frameworks related to nuclear fuel cycles in both countries were reviewed. The various advanced fuel cycle scenarios that have been considered in both countries were then surveyed and summarized. The working group established the working reference scenario for the future cooperative activity that corresponds to a fuel cycle scenario being considered both in Japan and the U.S. This working scenario involves transitioning from a once-through fuel cycle utilizing light water reactors to a one-pass uranium-plutonium fuel recycle in light water reactors to a combination of light water reactors and fast reactors with plutonium, uranium, and minor actinide recycle, ultimately concluding with multiple recycle passes primarily using fast reactors. Considering the scenario, current and future expected waste streams, treatment and inventory were discussed, and the relevant information was summarized. Second, the waste management/disposal system optimization was discussed. Repository system concepts were reviewed, repository design concepts for the various classifications of nuclear waste were summarized, and the factors to consider in repository design and optimization were then discussed. Japan is considering various alternatives and options for the geologic disposal facility and the framework for future analysis of repository concepts was discussed. Regarding the advanced waste and storage form development, waste form technologies developed in both countries were surveyed and compared. Potential collaboration areas and activities were next identified. Disposal system optimization processes and techniques were reviewed, and factors to consider in future repository design optimization activities were also discussed. Then the potential collaboration areas and activities related to the optimization problem were extracted.« less

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

DTIC Science & Technology

2013-12-01

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

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

NASA Astrophysics Data System (ADS)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

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

STAN, MARIUS; HECKER, SIEGFRIED S.

2007-02-07

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuelsmore » suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.« less

A study of passive safety features by utilizing intra-subassembly-equipped self-actuated shutdown mechanism for future large fast breeder reactors

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

Uto, N.; Niwa, H.; Ieda, Y.

1996-08-01

Passive prevention of core disruptive accidents (CDAs) is desired in terms of enhancement of safety for future fast breeder reactors. In addition, mitigation of CDA`s consequences should be required because mitigation measures have a potential of applying to all accidents, while prevention measures are prepared for specific accident initiators. In this paper, the Intra-Subassembly-equipped Self-Actuated Shutdown System (IS-SASS) , which is considered effective on passive prevention and mitigation of CDAs, is described. The IS-SASS is introduced in a fuel subassembly and consists of absorber materials at the top of the active core and an inner duct through which molten fuelmore » can be excluded out of the core. The determination of the appropriate number of the IS-SASS units, their arrangement in the core and their suitable structure are found to be suited to prevention and mitigation of CDAs for liquid metal-cooled large fast breeder reactors.« less

Biomass Conversion to Produce Hydrocarbon Liquid Fuel Via Hot-vapor Filtered Fast Pyrolysis and Catalytic Hydrotreating.

PubMed

Wang, Huamin; Elliott, Douglas C; French, Richard J; Deutch, Steve; Iisa, Kristiina

2016-12-25

Lignocellulosic biomass conversion to produce biofuels has received significant attention because of the quest for a replacement for fossil fuels. Among the various thermochemical and biochemical routes, fast pyrolysis followed by catalytic hydrotreating is considered to be a promising near-term opportunity. This paper reports on experimental methods used 1) at the National Renewable Energy Laboratory (NREL) for fast pyrolysis of lignocellulosic biomass to produce bio-oils in a fluidized-bed reactor and 2) at Pacific Northwest National Laboratory (PNNL) for catalytic hydrotreating of bio-oils in a two-stage, fixed-bed, continuous-flow catalytic reactor. The configurations of the reactor systems, the operating procedures, and the processing and analysis of feedstocks, bio-oils, and biofuels are described in detail in this paper. We also demonstrate hot-vapor filtration during fast pyrolysis to remove fine char particles and inorganic contaminants from bio-oil. Representative results showed successful conversion of biomass feedstocks to fuel-range hydrocarbon biofuels and, specifically, the effect of hot-vapor filtration on bio-oil production and upgrading. The protocols provided in this report could help to generate rigorous and reliable data for biomass pyrolysis and bio-oil hydrotreating research.

Biomass Conversion to Produce Hydrocarbon Liquid Fuel Via Hot-vapor Filtered Fast Pyrolysis and Catalytic Hydrotreating

PubMed Central

Wang, Huamin; Elliott, Douglas C.; French, Richard J.; Deutch, Steve; Iisa, Kristiina

2016-01-01

Lignocellulosic biomass conversion to produce biofuels has received significant attention because of the quest for a replacement for fossil fuels. Among the various thermochemical and biochemical routes, fast pyrolysis followed by catalytic hydrotreating is considered to be a promising near-term opportunity. This paper reports on experimental methods used 1) at the National Renewable Energy Laboratory (NREL) for fast pyrolysis of lignocellulosic biomass to produce bio-oils in a fluidized-bed reactor and 2) at Pacific Northwest National Laboratory (PNNL) for catalytic hydrotreating of bio-oils in a two-stage, fixed-bed, continuous-flow catalytic reactor. The configurations of the reactor systems, the operating procedures, and the processing and analysis of feedstocks, bio-oils, and biofuels are described in detail in this paper. We also demonstrate hot-vapor filtration during fast pyrolysis to remove fine char particles and inorganic contaminants from bio-oil. Representative results showed successful conversion of biomass feedstocks to fuel-range hydrocarbon biofuels and, specifically, the effect of hot-vapor filtration on bio-oil production and upgrading. The protocols provided in this report could help to generate rigorous and reliable data for biomass pyrolysis and bio-oil hydrotreating research. PMID:28060311

Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1973-01-01

The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

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

Kim, T. K.; Grandy, C.; Natesan, K.

The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treatedmore » separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s, and the remaining critical paths and R&D needs are generally related to the completion of qualification of fuel and structural materials, validation of reactor design codes and methods, and support of the licensing frameworks. The LFR’s technology is instead less-mature compared to the SFR’s, and will be at the engineering demonstration stage by the early 2030s. Key LFR technology development activities will focus on resolving remaining design challenges and demonstrating the viability of systems and components in the integral system, which will be done in parallel with addressing the gaps shared with SFR technology. The approach and timeline presented here assume that, for the first module demonstration, vendors would pursue a two-step licensing process based on 10CFR Part 50.« less

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Vented target elements for use in an isotope-production reactor. [LMFBR

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

Assemblies with both target and fuel pins in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

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

Temitope A. Taiwo; Samuel E. Bays; Abdullatif M. Yacout

2011-03-01

A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent ofmore » the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.« less

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

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

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

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

Fast, Parallel and Secure Cryptography Algorithm Using Lorenz's Attractor

NASA Astrophysics Data System (ADS)

Marco, Anderson Gonçalves; Martinez, Alexandre Souto; Bruno, Odemir Martinez

A novel cryptography method based on the Lorenz's attractor chaotic system is presented. The proposed algorithm is secure and fast, making it practical for general use. We introduce the chaotic operation mode, which provides an interaction among the password, message and a chaotic system. It ensures that the algorithm yields a secure codification, even if the nature of the chaotic system is known. The algorithm has been implemented in two versions: one sequential and slow and the other, parallel and fast. Our algorithm assures the integrity of the ciphertext (we know if it has been altered, which is not assured by traditional algorithms) and consequently its authenticity. Numerical experiments are presented, discussed and show the behavior of the method in terms of security and performance. The fast version of the algorithm has a performance comparable to AES, a popular cryptography program used commercially nowadays, but it is more secure, which makes it immediately suitable for general purpose cryptography applications. An internet page has been set up, which enables the readers to test the algorithm and also to try to break into the cipher.

ESS-FH: Enhanced Security Scheme for Fast Handover in Hierarchical Mobile IPv6

NASA Astrophysics Data System (ADS)

You, Ilsun; Lee, Jong-Hyouk; Sakurai, Kouichi; Hori, Yoshiaki

Fast Handover for Hierarchical Mobile IPv6 (F-HMIPv6) that combines advantages of Fast Handover for Mobile IPv6 (FMIPv6) and Hierarchical Mobile IPv6 (HMIPv6) achieves the superior performance in terms of handover latency and signaling overhead compared with previously developed mobility protocols. However, without being secured, F-HMIPv6 is vulnerable to various security threats. In 2007, Kang and Park proposed a security scheme, which is seamlessly integrated into F-HMIPv6. In this paper, we reveal that Kang-Park's scheme cannot defend against the Denial of Service (DoS) and redirect attacks while largely relying on the group key. Then, we propose an Enhanced Security Scheme for F-HMIPv6 (ESS-FH) that achieves the strong key exchange and the key independence as well as addresses the weaknesses of Kang-Park's scheme. More importantly, it enables fast handover between different MAP domains. The proposed scheme is formally verified based on BAN-logic, and its handover latency is analyzed and compared with that of Kang-Park's scheme.

Reactor Statics Module, RS-9: Multigroup Diffusion Program Using an Exponential Acceleration Technique.

ERIC Educational Resources Information Center

Macek, Victor C.

The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burnup for both slow neutron and fast neutron fission reactors. The last module, RS-9,…

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

HEDL FACILITIES CATALOG 400 AREA

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

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

77 FR 68162 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-15

... Water Reactors,'' Revision 2, and Regulatory Guide 1.79.1, ``Initial Test Program of Emergency Core... White Flint North building, 11555 Rockville Pike, Rockville, MD. After registering with security, please...

Safeguards and Security by Design (SSBD) for Small Modular Reactors (SMRs) through a Common Global Approach

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

Badwan, Faris M.; Demuth, Scott Francis; Miller, Michael Conrad

Small Modular Reactors (SMR) with power levels significantly less than the currently standard 1000 to 1600-MWe reactors have been proposed as a potential game changer for future nuclear power. SMRs may offer a simpler, more standardized, and safer modular design by using factory built and easily transportable components. Additionally, SMRs may be more easily built and operated in isolated locations, and may require smaller initial capital investment and shorter construction times. Because many SMRs designs are still conceptual and consequently not yet fixed, designers have a unique opportunity to incorporate updated design basis threats, emergency preparedness requirements, and then fullymore » integrate safety, physical security, and safeguards/material control and accounting (MC&A) designs. Integrating safety, physical security, and safeguards is often referred to as integrating the 3Ss, and early consideration of safeguards and security in the design is often referred to as safeguards and security by design (SSBD). This paper describes U.S./Russian collaborative efforts toward developing an internationally accepted common approach for implementing SSBD/3Ss for SMRs based upon domestic requirements, and international guidance and requirements. These collaborative efforts originated with the Nuclear Energy and Nuclear Security working group established under the U.S.-Russia Bilateral Presidential Commission during the 2009 Presidential Summit. Initial efforts have focused on review of U.S. and Russian domestic requirements for Security and MC&A, IAEA guidance for security and MC&A, and IAEA requirements for international safeguards. Additionally, example SMR design features that can enhance proliferation resistance and physical security have been collected from past work and reported here. The development of a U.S./Russian common approach for SSBD/3Ss should aid the designer of SMRs located anywhere in the world. More specifically, the application of this approach may lead to more proliferation resistant and physically secure design features for SMRs.« less

Fuel pins with both target and fuel pellets in an isotope-production reactor

DOEpatents

Cawley, W.E.; Omberg, R.P.

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

Actinide management with commercial fast reactors

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

Ohki, Shigeo

The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

Fast reactor core concepts to improve transmutation efficiency

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

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

Computational study: Reduction of iron corrosion in lead coolant of fast nuclear reactor

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

Arkundato, Artoto; Su'ud, Zaki; Abdullah, Mikrajuddin

2012-06-20

In this paper we report molecular dynamics simulation results of iron (cladding) corrosion in interaction with lead coolant of fast nuclear reactor. The goal of this work is to study effect of oxygen injection to the coolant to reduce iron corrosion. By evaluating diffusion coefficients, radial distribution functions, mean-square displacement curves and observation of crystal structure of iron before and after oxygen injection, we concluded that a significant reduction of corrosion can be achieved by issuing about 2% of oxygen atoms into lead coolant.

Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

NASA Astrophysics Data System (ADS)

Raj, Baldev; Rao, K. Bhanu Sankara

2009-04-01

The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

A Roadmap of Innovative Nuclear Energy System

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

Nuclear is a dense energy without CO2 emission. It can be used for more than 100,000 years using fast breeder reactors with uranium from the sea. However, it raises difficult problems associated with severe accidents, spent fuel waste and nuclear threats, which should be solved with acceptable costs. Some innovative reactors have attracted interest, and many designs have been proposed for small reactors. These reactors are considered much safer than conventional large reactors and have fewer technical obstructions. Breed-and-burn reactors have high potential to solve all inherent problems for peaceful use of nuclear energy. However, they have some technical problems with materials. A roadmap for innovative reactors is presented herein.

Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

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

Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

1997-04-01

Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop amore » variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.« less

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).

The basic features of a closed fuel cycle without fast reactors

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

Method and apparatus for obtaining enhanced production rate of thermal chemical reactions

DOEpatents

Tonkovich, Anna Lee Y.; Wang, Yong; Wegeng, Robert S.; Gao, Yufei

2003-09-09

Reactors and processes are disclosed that can utilize high heat fluxes to obtain fast, steady-state reaction rates. Porous catalysts used in conjunction with microchannel reactors to obtain high rates of heat transfer are also disclosed. Reactors and processes that utilize short contact times, high heat flux and low pressure drop are described. Improved methods of steam reforming are also provided.

BIOLOGICAL IRRADIATION FACILITY

DOEpatents

McCorkle, W.H.; Cern, H.S.

1962-04-24

A facility for irradiating biological specimens with neutrons is described. It includes a reactor wherein the core is off center in a reflector. A high-exposure room is located outside the reactor on the side nearest the core while a low-exposure room is located on the opposite side. Means for converting thermal neutrons to fast neutrons are movably disposed between the reactor core and the high and low-exposure rooms. (AEC)

Method and apparatus for obtaining enhanced production rate of thermal chemical reactions

DOEpatents

Tonkovich, Anna Lee Y [Pasco, WA; Wang, Yong [Richland, WA; Wegeng, Robert S [Richland, WA; Gao, Yufei [Kennewick, WA

2006-05-16

Reactors and processes are disclosed that can utilize high heat fluxes to obtain fast, steady-state reaction rates. Porous catalysts used in conjunction with microchannel reactors to obtain high rates of heat transfer are also disclosed. Reactors and processes that utilize short contact times, high heat flux and low pressure drop are described. Improved methods of steam reforming are also provided.

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

NUCLEAR FUEL COMPOSITION

DOEpatents

Spedding, F.H.; Wilhelm, H.A.

1960-05-31

A novel reactor composition for use in a self-sustaining fast nuclear reactor is described. More particularly, a fuel alloy comprising thorium and uranium-235 is de scribed, the uranium-235 existing in approximately the same amount that it is found in natural uranium, i.e., 1.4%.

Spectral unfolding of fast neutron energy distributions

NASA Astrophysics Data System (ADS)

Mosby, Michelle; Jackman, Kevin; Engle, Jonathan

2015-10-01

The characterization of the energy distribution of a neutron flux is difficult in experiments with constrained geometry where techniques such as time of flight cannot be used to resolve the distribution. The measurement of neutron fluxes in reactors, which often present similar challenges, has been accomplished using radioactivation foils as an indirect probe. Spectral unfolding codes use statistical methods to adjust MCNP predictions of neutron energy distributions using quantified radioactive residuals produced in these foils. We have applied a modification of this established neutron flux characterization technique to experimentally characterize the neutron flux in the critical assemblies at the Nevada National Security Site (NNSS) and the spallation neutron flux at the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). Results of the unfolding procedure are presented and compared with a priori MCNP predictions, and the implications for measurements using the neutron fluxes at these facilities are discussed.

76 FR 20377 - Applications and Amendments to Facility Operating Licenses Involving Proposed No Significant...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-04-12

... amendment requests approval of the Exelon Cyber Security Plan, provides an Implementation Schedule, and adds... require Exelon to fully implement and maintain in effect all provisions of the approved Cyber Security..., Revision 6, ``Cyber Security Plan for Nuclear Power Reactors.'' Basis for proposed no significant hazards...

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

2016-12-01

The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

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

Thompson, Logan C.; Ciesielski, Peter N.; Jarvis, Mark W.

Here, biomass particles can experience variable thermal conditions during fast pyrolysis due to differences in their size and morphology, and from local temperature variations within a reactor. These differences lead to increased heterogeneity of the chemical products obtained in the pyrolysis vapors and bio-oil. Here we present a simple, high-throughput method to investigate the thermal history experienced by large ensembles of particles during fast pyrolysis by imaging and quantitative image analysis. We present a correlation between the surface luminance (darkness) of the biochar particle and the highest temperature that it experienced during pyrolysis. Next, we apply this correlation to large,more » heterogeneous ensembles of char particles produced in a laminar entrained flow reactor (LEFR). The results are used to interpret the actual temperature distributions delivered by the reactor over a range of operating conditions.« less

Fluid-structure interaction in fast breeder reactors

NASA Astrophysics Data System (ADS)

Mitra, A. A.; Manik, D. N.; Chellapandi, P. A.

2004-05-01

A finite element model for the seismic analysis of a scaled down model of Fast breeder reactor (FBR) main vessel is proposed to be established. The reactor vessel, which is a large shell structure with a relatively thin wall, contains a large volume of sodium coolant. Therefore, the fluid structure interaction effects must be taken into account in the seismic design. As part of studying fluid-structure interaction, the fundamental frequency of vibration of a circular cylindrical shell partially filled with a liquid has been estimated using Rayleigh's method. The bulging and sloshing frequencies of the first four modes of the aforementioned system have been estimated using the Rayleigh-Ritz method. The finite element formulation of the axisymmetric fluid element with Fourier option (required due to seismic loading) is also presented.

Thermal baffle for fast-breeder reacton

DOEpatents

Rylatt, John A.

1977-01-01

A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel.

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

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

S. T. Khericha

2007-04-01

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed tomore » achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.« less

Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

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

Heidet, F.; Kim, T.; Grandy, C.

2012-07-30

Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium ismore » more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.« less

10 CFR 73.4 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

...: Document Control Desk, Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear Material Safety and Safeguards, or Director, Division of Security Policy... 10 Energy 2 2012-01-01 2012-01-01 false Communications. 73.4 Section 73.4 Energy NUCLEAR...

Biomass conversion to produce hydrocarbon liquid fuel via hot-vapor filtered fast pyrolysis and catalytic hydrotreating

DOE PAGES

Wang, Huamin; Elliott, Douglas C.; French, Richard J.; ...

2016-12-25

Lignocellulosic biomass conversion to produce biofuels has received significant attention because of the quest for a replacement for fossil fuels. Among the various thermochemical and biochemical routes, fast pyrolysis followed by catalytic hydrotreating is considered to be a promising near-term opportunity. This paper reports on experimental methods used 1) at the National Renewable Energy Laboratory (NREL) for fast pyrolysis of lignocellulosic biomass to produce bio-oils in a fluidized-bed reactor and 2) at Pacific Northwest National Laboratory (PNNL) for catalytic hydrotreating of bio-oils in a two-stage, fixed-bed, continuous-flow catalytic reactor. The configurations of the reactor systems, the operating procedures, and themore » processing and analysis of feedstocks, bio-oils, and biofuels are described in detail in this paper. We also demonstrate hot-vapor filtration during fast pyrolysis to remove fine char particles and inorganic contaminants from bio-oil. Representative results showed successful conversion of biomass feedstocks to fuel-range hydrocarbon biofuels and, specifically, the effect of hot-vapor filtration on bio-oil production and upgrading. As a result, the protocols provided in this report could help to generate rigorous and reliable data for biomass pyrolysis and bio-oil hydrotreating research.« less

Biomass conversion to produce hydrocarbon liquid fuel via hot-vapor filtered fast pyrolysis and catalytic hydrotreating

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

Wang, Huamin; Elliott, Douglas C.; French, Richard J.

Lignocellulosic biomass conversion to produce biofuels has received significant attention because of the quest for a replacement for fossil fuels. Among the various thermochemical and biochemical routes, fast pyrolysis followed by catalytic hydrotreating is considered to be a promising near-term opportunity. This paper reports on experimental methods used 1) at the National Renewable Energy Laboratory (NREL) for fast pyrolysis of lignocellulosic biomass to produce bio-oils in a fluidized-bed reactor and 2) at Pacific Northwest National Laboratory (PNNL) for catalytic hydrotreating of bio-oils in a two-stage, fixed-bed, continuous-flow catalytic reactor. The configurations of the reactor systems, the operating procedures, and themore » processing and analysis of feedstocks, bio-oils, and biofuels are described in detail in this paper. We also demonstrate hot-vapor filtration during fast pyrolysis to remove fine char particles and inorganic contaminants from bio-oil. Representative results showed successful conversion of biomass feedstocks to fuel-range hydrocarbon biofuels and, specifically, the effect of hot-vapor filtration on bio-oil production and upgrading. As a result, the protocols provided in this report could help to generate rigorous and reliable data for biomass pyrolysis and bio-oil hydrotreating research.« less

Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

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

Benoit, Timothy; Hlotke, John Daniel; Yacout, Abdellatif

2017-07-05

This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generatedmore » during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.« less

Hodoscope Cineradiography Of Nuclear Fuel Destruction Experiments

NASA Astrophysics Data System (ADS)

De Volpi, A.

1983-08-01

Nuclear reactor safety studies have applied cineradiographic techniques to achieve key information regarding the durability of fuel elements that are subjected to destructive transients in test reactors. Beginning with its development in 1963, the fast-neutron hodoscope has recorded data at the TREAT reactor in the United States of America. Consisting of a collimator instrumented with several hundred parallel channels of detectors and associated instrumentation, the hodoscope measures fuel motion that takes place within thick-walled steel test containers. Fuel movement is determined by detecting the emission of fast neutrons induced in the test capsule by bursts of the test reactor that last from 0.3 to 30 s. The system has been designed so as to achieve under certain typical conditions( horizontal) spatial resolution less than lmm, time resolution close to lms, mass resolution below 0.1 g, with adequate dynamic range and recording duration. A variety of imaging forms have been developed to display the results of processing and analyzing recorded data.*

Temperature Resistant Fiber Bragg Gratings for On-Line and Structural Health Monitoring of the Next-Generation of Nuclear Reactors.

PubMed

Laffont, Guillaume; Cotillard, Romain; Roussel, Nicolas; Desmarchelier, Rudy; Rougeault, Stéphane

2018-06-02

The harsh environment associated with the next generation of nuclear reactors is a great challenge facing all new sensing technologies to be deployed for on-line monitoring purposes and for the implantation of SHM methods. Sensors able to resist sustained periods at very high temperatures continuously as is the case within sodium-cooled fast reactors require specific developments and evaluations. Among the diversity of optical fiber sensing technologies, temperature resistant fiber Bragg gratings are increasingly being considered for the instrumentation of future nuclear power plants, especially for components exposed to high temperature and high radiation levels. Research programs are supporting the developments of optical fiber sensors under mixed high temperature and radiative environments leading to significant increase in term of maturity. This paper details the development of temperature-resistant wavelength-multiplexed fiber Bragg gratings for temperature and strain measurements and their characterization for on-line monitoring into the liquid sodium used as a coolant for the next generation of fast reactors.

High-Sensitivity Fast Neutron Detector KNK-2-8M

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Dovbysh, L. Ye.; Ovchinnikov, M. A.; Pikulina, G. N.; Drozdov, Yu. M.; Chuklyaev, S. V.; Pepyolyshev, Yu. N.

2017-12-01

The design of the fast neutron detector KNK-2-8M is outlined. The results of he detector study in the pulse counting mode with pulses from 238U nuclei fission in the radiator of the neutron-sensitive section and in the current mode with separation of functional section currents are presented. The possibilities of determination of the effective number of 238U nuclei in the radiator of the neutron-sensitive section are considered. The diagnostic capabilities of the detector in the counting mode are demonstrated, as exemplified by the analysis of reference data on characteristics of neutron fields in the BR-1 reactor hall. The diagnostic capabilities of the detector in the current mode are demonstrated, as exemplified by the results of measurements of 238U fission intensity in the power startup of the BR-K1 reactor in the fission pulse generation mode with delayed neutrons and the detector placed in the reactor cavity in conditions of large-scale variation of the reactor radiation fields.

The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

NASA Astrophysics Data System (ADS)

Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

2018-01-01

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

NASA Astrophysics Data System (ADS)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

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

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

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

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

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

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

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

Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

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

Graf, A.; Petrick, H.; Stutz, U.

2012-07-01

The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up intomore » small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and also foreign decommissioning projects can provide solutions with were unknown before. (authors)« less

Sustained Recycle in Light Water and Sodium-Cooled Reactors

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

Steven J. Piet; Samuel E. Bays; Michael A. Pope

2010-11-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in freshmore » fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.« less

Risk Management for Sodium Fast Reactors.

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

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

2015-01-01

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

Economics and Environmental Compatibility of Fusion Reactors —Its Analysis and Coming Issues— 4.Economic Effect of Fusion in Energy Market 4.2Various Externalities of Energy Systems and the Integrated Evaluation

NASA Astrophysics Data System (ADS)

Ito, Keishiro

The primacy of a nuclear fusion reactor in a competitive energy market remarkably depends on to what extent the reactor contributes to reduce the externalities of energy. The reduction effects are classified into two effects, which have quite dissimilar characteristics. One is an effect of environmental dimensions. The other is related to energy security. In this study I took up the results of EC's Extern Eproject studies as are presentative example of the former effect. Concerning the latter effect, I clarified the fundamental characteristics of externalities related to energy security and the conceptual framework for the purpose of evaluation. In the socio-economical evaluation of research into and development investments in nuclear fusions reactors, the public will require the development of integrated evaluation systems to support the cost-effect analysis of how well the reduction effects of externalities have been integrated with the effects of technological innovation, learning, spillover, etc.

Nuclear reactor pressure vessel support system

DOEpatents

Sepelak, George R.

1978-01-01

A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

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

Lucia M. Petkovic; Daniel M. Ginosar

Since the year 2000, the United States Patent and Trademark Office (USPTO) has granted a dozen patents for inventions related to methane dehydroaromatization processes. One of them was granted to UOP LLC (Des Plaines). It relates to a catalyst composition and preparation method. Two patents were granted to Conoco Phillips Company (Houston, TX). One was aimed at securing a process and operating conditions for methane aromatization. The other was aimed at securing a process that may be integrated with separation of wellhead fluids and blending of the aromatics produced from the gas with the crude. Nine patents were granted tomore » ExxonMobil Chemical Patents Inc. (Houston, TX). Most of these were aimed at securing a dehydroaromatization process where methane-containing feedstock moves counter currently to a particulate catalyst. The coked catalyst is heated or regenerated either in the reactor, by cyclic operation, or in annex equipment, and returned to the reactor. The reactor effluent stream may be separated in its main components and used or recycled as needed. A brief summary of those inventions is presented in this review.« less

Analysis of Trace VX in Acidified VX Hydrolysate Samples

DTIC Science & Technology

2009-07-01

SUBJECT TERMS Mass spectrometry Gas chromatography VX hydrolysate Energetics Blue Grass VX reformation BGCAPP 16. SECURITY CLASSIFICATION OF: 17...SCWO ( supercritical water oxidation) reactors. Prior to feeding the blended hydrolysate mixture from the SCWO blend tank to the SCWO reactors, chloride...transported as fluid in the reactor under the SCWO processing conditions. Current design calls for adding these elements as 35% HCI, 93% H2SO4 and

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

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

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

Integral Fast Reactor fuel pin processor

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

Levinskas, D.

1993-01-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Integral Fast Reactor fuel pin processor

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

Levinskas, D.

1993-03-01

This report discusses the pin processor which receives metal alloy pins cast from recycled Integral Fast Reactor (IFR) fuel and prepares them for assembly into new IFR fuel elements. Either full length as-cast or precut pins are fed to the machine from a magazine, cut if necessary, and measured for length, weight, diameter and deviation from straightness. Accepted pins are loaded into cladding jackets located in a magazine, while rejects and cutting scraps are separated into trays. The magazines, trays, and the individual modules that perform the different machine functions are assembled and removed using remote manipulators and master-slaves.

Dynamic Computer Model of a Stirling Space Nuclear Power System

DTIC Science & Technology

2006-05-04

diagram of electric propulsion…………………………………. 17 Figure 2-1. General NEP structure……………………………………………………….20 Figure 2-2. Fission of uranium -235...Figure 2-1. General NEP structure. [20] 21 Figure 2-2. Fission of uranium -235. In a fast reactor, the average number of neutrons...that is modeled for this project is a 600 kW(t) fast fission reactor consisting of uranium nitride fuel and sodium potassium coolant. Its dynamic

Nuclear and Physical Properties of Dielectrics under Neutron Irradiation in Fast (BN-600) and Fusion (DEMO-S) Reactors

NASA Astrophysics Data System (ADS)

Blokhin, D. A.; Chernov, V. M.; Blokhin, A. I.

2017-12-01

Nuclear and physical properties (activation and transmutation of elements) of BN and Al2O3 dielectric materials subjected to neutron irradiation for up to 5 years in Russian fast (BN-600) and fusion (DEMO-S) reactors were calculated using the ACDAM-2.0 software complex for different post-irradiation cooling times (up to 10 years). Analytical relations were derived for the calculated quantities. The results may be used in the analysis of properties of irradiated dielectric materials and may help establish the rules for safe handling of these materials.

Austenitic alloy and reactor components made thereof

DOEpatents

Bates, John F.; Brager, Howard R.; Korenko, Michael K.

1986-01-01

An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

Prospective scenarios of nuclear energy evolution over the 21. century

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

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

SLSF in-reactor local fault safety experiment P4. Final report

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

Thompson, D. H.; Holland, J. W.; Braid, T. H.

The Sodium Loop Safety Facility (SLSF), a major facility in the US fast-reactor safety program, has been used to simulate a variety of sodium-cooled fast reactor accidents. SLSF experiment P4 was conducted to investigate the behavior of a "worse-than-case" local fault configuration. Objectives of this experiment were to eject molten fuel into a 37-pin bundle of full-length Fast-Test-Reactor-type fuel pins form heat-generating fuel canisters, to characterize the severity of any molten fuel-coolant interaction, and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors to prevent fuel failure propagation. The designmore » goal for molten fuel release was 10 to 30 g. Explusion of molten fuel from fuel canisters caused failure of adjacent pins and a partial flow channel blockage in the fuel bundle during full-power operation. Molten fuel and fuel debris also lodged against the inner surface of the test subassembly hex-can wall. The total fuel disruption of 310 g evaluated from posttest examination data was in excellent agreement with results from the SLSF delayed neutron detection system, but exceeded the target molten fuel release by an order of magnitude. This report contains a summary description of the SLSF in-reactor loop and support systems and the experiment operations. results of the detailed macro- and microexamination of disrupted fuel and metal and results from the analysis of the on-line experimental data are described, as are the interpretations and conclusions drawn from the posttest evaluations. 60 refs., 74 figs.« less

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

NASA Astrophysics Data System (ADS)

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

2017-07-01

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

Hanford Works monthly report, December 1950

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

Prout, G.R.

1951-01-22

This is a progress report of the production reactors on the Hanford Reservation for the month of December 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, April 1952

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

Prout, G.R.

1952-05-20

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, August 1950

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

Prout, G.R.

1950-09-18

This is a progress report of the production reactors on the Hanford Reservation for the month of August 1950. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, May 1951

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

Prout, G.R.

1951-06-21

This is a progress report of the production reactors on the Hanford Reservation for the month of May 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, December 1951

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

Prout, G.R.

1952-01-22

This is a progress report of the production reactors on the Hanford Reservation for the month of December 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, March 1951

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

Prout, G.R.

1951-04-20

This is a progress report of the production reactors on the Hanford Reservation for the month of March 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1951

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, March 1952

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

Prout, G.R.

1952-04-18

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1952

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

Prout, G.R.

1952-08-15

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford works monthly report, September 1951

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, January 1952

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

Prout, G.R.

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1952. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2014 CFR

2014-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

Hanford Works monthly report, August 1951

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

Not Available

1951-09-24

This is a progress report of the production reactors on the Hanford Reservation for the month of August 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, July 1950

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

Prout, G.R.

1950-08-18

This is a progress report of the production reactors on the Hanford Reservation for the month of July 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, November 1951

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

Prout, G.R.

1951-12-21

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, October 1950

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

Prout, G.R.

1950-11-20

This is a progress report of the production reactors on the Hanford Reservation for the month of October 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, September 1950

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

Prout, G.R.

1950-10-20

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1950. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Works monthly report, November 1950

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

Prout, G.R.

1950-12-20

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1950. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

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

Lubina, A. S., E-mail: lubina-as@nrcki.ru; Subbotin, A. S.; Sedov, A. A.

2016-12-15

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The resultsmore » of the hydrodynamics and heat transfer calculations have been analyzed.« less

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

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

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

2015-12-15

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

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

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

NASA Astrophysics Data System (ADS)

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

2014-09-01

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

Development work for a borax internal core-catcher for a gas-cooled fast reactor

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

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

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

Prokopowicz, R.; Pytel, K.; Dorosz, M.

2015-07-01

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction aremore » essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)« less

Interim status report on lead-cooled fast reactor (LFR) research and development.

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

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

2008-03-31

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

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Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

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

Perret, G.; Pattupara, R. M.; Girardin, G.

2012-07-01

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fastmore » Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)« less

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

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

Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

Exploding the myths about the fast breeder reactor

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

Burns, S.

1979-01-01

This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

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

Not Available

The SPS Concept Development and Evaluation Program includes a comparative assessment. An early first step in the assessment process is the selection and characterization of alternative technologies. This document describes the cost and performance (i.e., technical and environmental) characteristics of six central station energy alternatives: (1) conventional coal-fired powerplant; (2) conventional light water reactor (LWR); (3) combined cycle powerplant with low-Btu gasifiers; (4) liquid metal fast breeder reactor (LMFBR); (5) photovoltaic system without storage; and (6) fusion reactor.

Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

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

Hellesen, C.; Grape, S.; Haakanson, A.

2013-07-01

Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

Thermal barrier and support for nuclear reactor fuel core

DOEpatents

Betts, Jr., William S.; Pickering, J. Larry; Black, William E.

1987-01-01

A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

Results of the International Energy Agency Round Robin on Fast Pyrolysis Bio-oil Production

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

Elliott, Douglas C.; Meier, Dietrich; Oasmaa, Anja

An international round robin study of the production of fast pyrolysis bio-oil was undertaken. Fifteen institutions in six countries contributed. Three biomass samples were distributed to the laboratories for processing in fast pyrolysis reactors. Samples of the bio-oil produced were transported to a central analytical laboratory for analysis. The round robin was focused on validating the pyrolysis community understanding of production of fast pyrolysis bio-oil by providing a common feedstock for bio-oil preparation. The round robin included: •distribution of 3 feedstock samples from a common source to each participating laboratory; •preparation of fast pyrolysis bio-oil in each laboratory with themore » 3 feedstocks provided; •return of the 3 bio-oil products (minimum 500 ml) with operational description to a central analytical laboratory for bio-oil property determination. The analyses of interest were: density, viscosity, dissolved water, filterable solids, CHN, S, trace element analysis, ash, total acid number, pyrolytic lignin, and accelerated aging of bio-oil. In addition, an effort was made to compare the bio-oil components to the products of analytical pyrolysis through GC/MS analysis. The results showed that clear differences can occur in fast pyrolysis bio-oil properties by applying different reactor technologies or configurations. The comparison to analytical pyrolysis method suggested that Py-GC/MS could serve as a rapid screening method for bio-oil composition when produced in fluid-bed reactors. Furthermore, hot vapor filtration generally resulted in the most favorable bio-oil product, with respect to water, solids, viscosity, and total acid number. These results can be helpful in understanding the variation in bio-oil production methods and their effects on bio-oil product composition.« less

Hanford Atomic Products Operation monthly report, April 1954

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

McCune, F.K.

1954-05-21

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1954. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, March 1953

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

McCune, F.K.

1953-04-22

This is a progress report of the production reactors on the Hanford Reservation for the month of March 1953. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, April 1953

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

McCune, F.K.

1953-05-20

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

Hanford Atomic Products Operation monthly report, January 1954

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

McCune, F.K.

1954-02-25

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1954. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes the accomplishments and employee relations for that month.

In situ TEM and synchrotron characterization of U–10Mo thin specimen annealed at the fast reactor temperature regime

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

Yun, Di, E-mail: diyun1979@xjtu.edu.cn; Xi'an Jiao Tong University, 28 Xian Ning West Road, Xi'an 710049; Mo, Kun

2015-12-15

U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratorymore » (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen. - Highlights: • TEM annealing experiments were performed in situ at the IVEM facility up to fast reactor temperature. • At 540 °C, the U-10Mo specimen underwent a slow microstructure transition where nano-sized grains were observed to emerge. • UO{sub 2} phase exists at the thin area of the as-annealed specimen whereas U-10Mo γ phase dominated at the thicker part. • Bcc γ U-10Mo recrystallized to become nano-meter sized crystallites near the specimen surface. • A separateannealing experiment was conducted with a FIB processed specimen where similar transition occurred at a lower temperature of 460 °C with a faster rate.« less

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.

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

ADS Model in the TIRELIRE-STRATEGIE Fuel Cycle Simulation Code Application to Minor Actinides Transmutation Studies

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

Garzenne, Claude; Massara, Simone; Tetart, Philippe

2006-07-01

Accelerator Driven Systems offer the advantage, thanks to the core sub-criticality, to burn highly radioactive elements such as americium and curium in a dedicated stratum, and then to avoid polluting with these elements the main part of the nuclear fleet, which is optimized for electricity production. This paper presents firstly the ADS model implemented in the fuel cycle simulation code TIRELIRE-STRATEGIE that we developed at EDF R and D Division for nuclear power scenario studies. Then we show and comment the results of TIRELIRE-STRATEGIE calculation of a transition scenario between the current French nuclear fleet, and a fast reactor fleetmore » entirely deployed towards the end of the 21. century, consistently with the EDF prospective view, with 3 options for the minor actinides management:1) vitrified with fission products to be sent to the final disposal; 2) extracted together with plutonium from the spent fuel to be transmuted in Generation IV fast reactors; 3) eventually extracted separately from plutonium to be incinerated in a ADSs double stratum. The comparison of nuclear fuel cycle material fluxes and inventories between these options shows that ADSs are not more efficient than critical fast reactors for reducing the high level waste radio-toxicity; that minor actinides inventory and fluxes in the fuel cycle are more than twice as high in case of a double ADSs stratum than in case of minor actinides transmutation in Generation IV FBRs; and that about fourteen 400 MWth ADS are necessary to incinerate minor actinides issued from a 60 GWe Generation IV fast reactor fleet, corresponding to the current French nuclear fleet installed power. (authors)« less

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Ogawa, Y.; Goto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Someya, Y.

2008-05-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions

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

Ruan, Guihua, E-mail: guihuaruan@hotmail.com; Guangxi Collaborative Innovation Center for Water Pollution Control and Water Safety in Karst Area, Guilin University of Technology, Guilin 541004; Wu, Zhenwei

A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of N{sub α}-benzoyl-L-arginine ethyl ester to N{sub α}-benzoyl-L-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast andmore » easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. - Graphical abstract: Schematic illustration of preparation of hypercrosslinking polyHIPE immobilized enzyme reactor for on-column protein digestion. - Highlights: • A reactor was prepared and used for enzyme immobilization and continuous on-column protein digestion. • The new polyHIPE IMER was quite suit for protein digestion with good properties. • On-column digestion revealed that the IMER was easy regenerated by HCl without any structure destruction.« less

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

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

Duo, J. I.

2011-07-01

Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reducedmore » computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)« less

Experimental Measurements at the MASURCA Facility

NASA Astrophysics Data System (ADS)

Assal, W.; Bosq, J. C.; Mellier, F.

2012-12-01

Dedicated to the neutronics studies of fast and semi-fast reactor lattices, MASURCA (meaning “mock-up facility for fast breeder reactor studies at CADARACHE”) is an airflow cooled fast reactor operating at a maximum power of 5 kW playing an important role in the CEA research activities. At this facility, a lot of neutron integral experimental programs were undertaken. The purpose of this poster is to show a panorama of the facility from this experimental measurement point of view. A hint at the forthcoming refurbishment will be included. These programs include various experimental measurements (reactivity, distributions of fluxes, reaction rates), performed essentially with fission chambers, in accordance with different methods (noise methods, radial or axial traverses, rod drops) and involving several devices systems (monitors, fission chambers, amplifiers, power supplies, data acquisition systems ...). For this purpose are implemented electronics modules to shape the signals sent from the detectors in various mode (fluctuation, pulse, current). All the electric and electronic devices needed for these measurements and the relating wiring will be fully explained through comprehensive layouts. Data acquired during counting performed at the time of startup phase or rod drops are analyzed by the mean of a Neutronic Measurement Treatment (TMN in French) programmed on the basis of the MATLAB software. This toolbox gives the opportunity of data files management, reactivity valuation from neutronics measurements and transient or divergence simulation at zero power. Particular TMN using at MASURCA will be presented.

Radiolysis of water at elevated temperatures—III. Simulation of radiolytic products at 25 and 250°C under the irradiation with γ-rays and fast neutrons

NASA Astrophysics Data System (ADS)

Sunaryo, Geni R.; Katsumura, Yosuke; Ishigure, Kenkichi

1995-05-01

The G-values of water decomposition products under the irradiations with γ-rays and fast neutrons up to 250°C have been determined in previous studies. In order to clarify the characteristics of the determined G-values, computer simulations under the simplified conditions in nuclear reactors have been carried out. The recent G-values for γ-radiolysis reported by Elliot, Chenier and Quellete [(1990) Can. J. Chem.68, 712; (1993) J. Chem. Soc. Faraday Trans.89, 1193], Kent and Sims [(1992) Water Chemistry of Nuclear Reactor Systems 6, p. 153. BNES, London], and Sunaryo, Katsumura, Shirai, Hiroishi and Ishigure [(1994) Radiat. Phys. Chem.44, 273] and Sunaryo, Katsumura, Hiroishi and Ishigure [(1995) Radiat. Phys. Chem.45, 131] are almost equivalent from the point of simulations. On the contrary, G-values for fast neutron radiolysis give a significant influence to the result, which arises from the higher molecular yields and smaller radical yields of water decomposition in fast neutron radiolysis, and it has been revealed that the dose evaluation in the reactor is inevitably important. In addition, it was pointed out by the simulations that reverse reactions for H 2+ .OH→ .H+H 2O and e aq-+H +→ .H, be neglected at room temperature, become important at higher temperatures.

The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

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

De Bruyn, D.; Engelen, J.; Ortega, A.

MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of threemore » years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)« less

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

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

Ariani, Menik; Su'ud, Zaki; Waris, Abdul

2012-06-06

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it ismore » shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.« less

78 FR 32476 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Digital I&C...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-30

..., June 4, 2013--1:00 p.m. Until 5:00 p.m. The Subcommittee will review and discuss all cyber security... Flint North building, 11555 Rockville Pike, Rockville, MD. After registering with security, please...

The software-defined fast post-processing for GEM soft x-ray diagnostics in the Tungsten Environment in Steady-state Tokamak thermal fusion reactor

NASA Astrophysics Data System (ADS)

Krawczyk, Rafał Dominik; Czarski, Tomasz; Linczuk, Paweł; Wojeński, Andrzej; Kolasiński, Piotr; GÄ ska, Michał; Chernyshova, Maryna; Mazon, Didier; Jardin, Axel; Malard, Philippe; Poźniak, Krzysztof; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol

2018-06-01

This article presents a novel software-defined server-based solutions that were introduced in the fast, real-time computation systems for soft X-ray diagnostics for the WEST (Tungsten Environment in Steady-state Tokamak) reactor in Cadarache, France. The objective of the research was to provide a fast processing of data at high throughput and with low latencies for investigating the interplay between the particle transport and magnetohydrodynamic activity. The long-term objective is to implement in the future a fast feedback signal in the reactor control mechanisms to sustain the fusion reaction. The implemented electronic measurement device is anticipated to be deployed in the WEST. A standalone software-defined computation engine was designed to handle data collected at high rates in the server back-end of the system. Signals are obtained from the front-end field-programmable gate array mezzanine cards that acquire and perform a selection from the gas electron multiplier detector. A fast, authorial library for plasma diagnostics was written in C++. It originated from reference offline MATLAB implementations. They were redesigned for runtime analysis during the experiment in the novel online modes of operation. The implementation allowed the benchmarking, evaluation, and optimization of plasma processing algorithms with the possibility to check the consistency with reference computations written in MATLAB. The back-end software and hardware architecture are presented with data evaluation mechanisms. The online modes of operation for the WEST are discussed. The results concerning the performance of the processing and the introduced functionality are presented.

An intrinsically safe facility for forefront research and training on nuclear technologies

NASA Astrophysics Data System (ADS)

Mansani, L.; Monti, S.; Ricco, G.; Ricotti, M.

2014-04-01

In this short paper the motivations for the development of fast spectrum lead-cooled reactors are briefly summarized. In particular the importance of subcritical research reactors, like the one described in this Focus Point, for the investigation of various scientifical and technological aspects and the training of students, is discussed.

Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

DOEpatents

Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

1977-01-01

A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

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

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Calculation of energetic characteristics of C-14 emitted from Beloyarsk nuclear power plant plume with fast neutron reactor

NASA Astrophysics Data System (ADS)

Kolotkov, Gennady A.; Penin, Sergei

2017-11-01

The paper examines an update of comparative analysis of radionuclides released into the atmosphere from Beloyarsk nuclear power plant with fast-neutron reactor for nine years in a row, from 2008 to 2016. It has been shown that the main radionuclides throw out into the atmosphere from Beloyarsk nuclear power plant are beta-active radionuclides. Based on data releases of the RPA "Typhoon", it has been conclude that radiation situation become worse insignificantly; beside on the new reactor BN-800 was put in operation in 2016. Using Spencer-Fano's equation, it was carried out the summary spectrum of emitted radionuclides. On example of Beloyarsk nuclear power plant, it was considered a question about ability of remote detection of raised radioactivity in the atmospheric radioactive plume. It has been shown that it possible to detect raised radioactivity in the emission plume from Beloyarsk nuclear power plant.

Catalyst Residence Time Distributions in Riser Reactors for Catalytic Fast Pyrolysis. Part 2: Pilot-Scale Simulations and Operational Parameter Study

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

Foust, Thomas D.; Ziegler, Jack L.; Pannala, Sreekanth

2017-02-21

Here, wsing the validated simulation model developed in part one of this study for biomass catalytic fast pyrolysis (CFP), we assess the functional utility of using this validated model to assist in the development of CFP processes in fluidized catalytic cracking (FCC) reactors to a commercially viable state. Specifically, we examine the effects of mass flow rates, boundary conditions (BCs), pyrolysis vapor molecular weight variation, and the impact of the chemical cracking kinetics on the catalyst residence times. The factors that had the largest impact on the catalyst residence time included the feed stock molecular weight and the degree ofmore » chemical cracking as controlled by the catalyst activity. Lastly, because FCC reactors have primarily been developed and utilized for petroleum cracking, we perform a comparison analysis of CFP with petroleum and show the operating regimes are fundamentally different.« less

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Elastic and inelastic neutron scattering cross sections for fission reactor applications

NASA Astrophysics Data System (ADS)

Hicks, S. F.; Chakraborty, A.; Combs, B.; Crider, B. P.; Downes, L.; Girgis, J.; Kersting, L. J.; Kumar, A.; Lueck, C. J.; McDonough, P. J.; McEllistrem, M. T.; Peters, E. E.; Prados-Estevz, F. M.; Schniederjan, J.; Sidwell, L.; Sigillito, A. J.; Vanhoy, J. R.; Watts, D.; Yates, S. W.

2013-04-01

Nuclear data important for the design and development of the next generation of light-water reactors and future fast reactors include neutron elastic and inelastic scattering cross sections on important structural materials, such as Fe, and on coolant materials, such as Na. These reaction probabilities are needed since neutron reactions impact fuel performance during irradiations and the overall efficiency of reactors. While neutron scattering cross sections from these materials are available for certain incident neutron energies, the fast neutron region, particularly above 2 MeV, has large gaps for which no measurements exist, or the existing uncertainties are large. Measurements have been made at the University of Kentucky Accelerator Laboratory to measure neutron scattering cross sections on both Fe and Na in the region where these gaps occur and to reduce the uncertainties on scattering from the ground state and first excited state of these nuclei. Results from measurements on Fe at incident neutron energies between 2 and 4 MeV will be presented and comparisons will be made to model calculations available from data evaluators.

High-sensitivity fast neutron detector KNK-2-7M

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Dovbysh, L. Ye.; Ovchinnikov, M. A.

2015-12-15

The construction of the fast neutron detector KNK-2-7M is briefly described. The results of the study of the detector in the pulse-counting mode are given for the fissions of {sup 237}Np nuclei in the radiator of the neutron-sensitive section and in the current mode with the separation of sectional currents of functional sections. The possibilities of determining the effective number of {sup 237}Np nuclei in the radiator of the neutronsensitive section are considered. The diagnostic possibilities of the detector in the counting mode are shown by example of the analysis of the reference data from the neutron-field characteristics in themore » working hall of the BR-K1 reactor. The diagnostic possibilities of the detector in the current operating mode are shown by example of the results of measuring the {sup 237}Np-fission intensity in the BR-K1 reactor power start-ups implemented in the mode of fission-pulse generation on delayed neutrons at the detector arrangement inside the reactor core cavity under conditions of a wide variation of the reactor radiation field.« less

Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

NASA Technical Reports Server (NTRS)

Hsieh, T.-M.; Koenig, D. R.

1977-01-01

Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

A low power ADS for transmutation studies in fast systems

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-12-01

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

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

Saleh, Tarik A.; Maloy, Stuart Andrew; Aydogan, Eda

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

A mechanism for proven technology foresight for emerging fast reactor designs and concepts

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

Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my

The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all ofmore » the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.« less

First overpower tests of metallic IFR [Integral Fast Reactor] fuel in TREAT [Transient Reactor Test Facility]: Data and analysis from tests M5, M6, and M7

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

Bauer, T. H.; Robinson, W. R.; Holland, J. W.

1989-12-01

Results and analyses of margin to cladding failure and pre-failure axial expansion of metallic fuel are reported for TREAT in-pile transient overpower tests M5--M7. These are the first such tests on reference binary and ternary alloy fuel of the Integral Fast Reactor (IFR) concept with burnup ranging from 1 to 10 at. %. In all cases, test fuel was subjected to an exponential power rise on an 8 s period until either incipient or actual cladding failure was achieved. Objectives, designs and methods are described with emphasis on developments unique to metal fuel safety testing. The resulting database for claddingmore » failure threshold and prefailure fuel expansion is presented. The nature of the observed cladding failure and resultant fuel dispersals is described. Simple models of cladding failures and pre-failure axial expansions are described and compared with experimental results. Reported results include: temperature, flow, and pressure data from test instrumentation; fuel motion diagnostic data principally from the fast neutron hodoscope; and test remains described from both destructive and non-destructive post-test examination. 24 refs., 144 figs., 17 tabs.« less

Small Modular Reactors: The Army’s Secure Source of Energy?

DTIC Science & Technology

2012-03-21

significant advantages of SMRs is the minimal amount of carbon dioxide (greenhouse gases) that is released in conjunction with the lifecycle operations...moderator in these reactors as well as the cooling agent and the means by which heat is removed to produce steam for turning the turbines of the...separate water system to generate steam to turn a turbine which then produces electricity. In the second type of light water reactors, the boiling water

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

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

Bruce G. Schnitzler; Stanley K. Borowski

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified asmore » the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.« less

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Schnitzler, Bruce G.; Borowski, Stanley K.

2012-01-01

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

Reducing Actinide Production Using Inert Matrix Fuels

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

Deinert, Mark

2017-08-23

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessingmore » that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.« less

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

PubMed

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

2014-01-01

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

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

PubMed Central

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

2014-01-01

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

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

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

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

DOE PAGES

Mamivand, Mahmood; Yang, Ying; Busby, Jeremy T.; ...

2017-03-11

The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300–400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 –7 dpa/s and 390 °C) and thenmore » use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 –8 dpa/s and 275 °C). The model accurately predicts the γ' (Ni 3Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γ' for typical CW 316 SS (with 0.054 wt% carbon) under LWR extended life conditions. Finally, this work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.« less

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...

SPERTI Reactor Pit Building (PER605) from contrasting direction as photo ...

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

SPERT-I Reactor Pit Building (PER-605) from contrasting direction as photo above (ID-33-F-32). Note Guard House door, security fencing around facility. Photographer: R.G. Larsen. Date: July 22, 1955. INEEL negative no. 55-1702. - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. CONTROL PANEL SUPPLIES STATUS INDICATORS. CARD ...

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

PROCESS WATER BUILDING, TRA-605. CONTROL PANEL SUPPLIES STATUS INDICATORS. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION ON ORIGINAL NEGATIVE. INL NEGATIVE NO. 4219. Unknown Photographer, 2/13/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

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

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

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

DOE PAGES

Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; ...

2015-07-17

Building new fission reactors in the United States presents many technical and regulatory challenges. Chief among the technical challenges is the need to share and present results from new high- fidelity, high- performance simulations in an easily consumable way. In light of the modern multi-scale, multi-physics simulations can generate petabytes of data, this will require the development of new techniques and methods to reduce the data to familiar quantities of interest with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately availablemore » in the community and need to be developed. Our paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It enables easy qualitative and quantitative comparisons between simulation results with a graphical user interface and cross-platform, multi-language input- output libraries for use by developers to work with the data. One example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented along with a detailed discussion of the system s requirements and design.« less

Characterization of alternative electric generation technologies for the SPS comparative assessment. Volume 1: Summary of central station technologies

NASA Astrophysics Data System (ADS)

1980-08-01

The technologies selected for the detailed characterization were: solar technology; terrestrial photovoltaic (200 MWe); coal technologies; conventional high sulfur coal combustion with advanced fine gas desulfurization (1250 MWe), and open cycle gas turbine combined cycle plant with low Btu gasifier (1250 MWe); and nuclear technologies: conventional light water reactor (1250 MWe), liquid metal fast breeder reactor (1250 MWe), and magnetic fusion reactor (1320 MWe). A brief technical summary of each power plant design is given.

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

A Novel Fast and Secure Approach for Voice Encryption Based on DNA Computing

NASA Astrophysics Data System (ADS)

Kakaei Kate, Hamidreza; Razmara, Jafar; Isazadeh, Ayaz

2018-06-01

Today, in the world of information communication, voice information has a particular importance. One way to preserve voice data from attacks is voice encryption. The encryption algorithms use various techniques such as hashing, chaotic, mixing, and many others. In this paper, an algorithm is proposed for voice encryption based on three different schemes to increase flexibility and strength of the algorithm. The proposed algorithm uses an innovative encoding scheme, the DNA encryption technique and a permutation function to provide a secure and fast solution for voice encryption. The algorithm is evaluated based on various measures including signal to noise ratio, peak signal to noise ratio, correlation coefficient, signal similarity and signal frequency content. The results demonstrate applicability of the proposed method in secure and fast encryption of voice files

Power monitoring in space nuclear reactors using silicon carbide radiation detectors

NASA Technical Reports Server (NTRS)

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

2005-01-01

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

Decision Model for U.S.- Mexico Border Security Measures

DTIC Science & Technology

2017-09-01

and money assigned to border security investments. 14. SUBJECT TERMS Department of Homeland Security (DHS), border security, U.S.–Mexico border...and money assigned to border security investments. vi THIS PAGE INTENTIONALLY LEFT BLANK vii TABLE OF CONTENTS I. INTRODUCTION...FAA Federal Aviation Administration FAMS Federal Air Marshals Service FAST Free and Secure Trade GSA General Services Administration HIR Human

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

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

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Fast particles in a steady-state compact FNS and compact ST reactor

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Nicolai, A.; Buxton, P.

2014-10-01

This paper presents results of studies of fast particles (ions and alpha particles) in a steady-state compact fusion neutron source (CFNS) and a compact spherical tokamak (ST) reactor with Monte-Carlo and Fokker-Planck codes. Full-orbit simulations of fast particle physics indicate that a compact high field ST can be optimized for energy production by a reduction of the necessary (for the alpha containment) plasma current compared with predictions made using simple analytic expressions, or using guiding centre approximation in a numerical code. Alpha particle losses may result in significant heating and erosion of the first wall, so such losses for an ST pilot plant have been calculated and total and peak wall loads dependence on the plasma current has been studied. The problem of dilution has been investigated and results for compact and big size devices are compared.

Pulse combustion reactor as a fast and scalable synthetic method for preparation of Li-ion cathode materials

NASA Astrophysics Data System (ADS)

Križan, Gregor; Križan, Janez; Dominko, Robert; Gaberšček, Miran

2017-09-01

In this work a novel pulse combustion reactor method for preparation of Li-ion cathode materials is introduced. Its advantages and potential challenges are demonstrated on two widely studied cathode materials, LiFePO4/C and Li-rich NMC. By exploiting the nature of efficiency of pulse combustion we have successfully established a slightly reductive or oxidative environment necessary for synthesis. As a whole, the proposed method is fast, environmentally friendly and easy to scale. An important advantage of the proposed method is that it preferentially yields small-sized powders (in the nanometric range) at a fast production rate of 2 s. A potential disadvantage is the relatively high degree of disorder of synthesized active material which however can be removed using a post-annealing step. This additional step allows a further tuning of materials morphology as shown and commented in some detail.

123. ARAI Substation (ARA726) plan, elevation, security fence details, and ...

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

123. ARA-I Substation (ARA-726) plan, elevation, security fence details, and sections. Norman Engineering Company 961-area/SF-726-E-1. Date: January 1959. Ineel index code no. 068-0726-10-613-102778. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

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

NUCLEAR REACTORS

DOEpatents

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

1961-12-01

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

Heat insulating system for a fast reactor shield slab

DOEpatents

Kotora Jr., James; Groh, Edward F.; Kann, William J.; Burelbach, James P.

1986-04-01

Improved thermal insulation for a nuclear reactor deck comprising many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

Heat insulating system for a fast reactor shield slab

DOEpatents

Kotora, Jr., James; Groh, Edward F.; Kann, William J.; Burelbach, James P.

1986-01-01

Improved thermal insulation for a nuclear reactor deck comprising many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

Heat insulating system for a fast reactor shield slab

DOEpatents

Kotora, J. Jr.; Groh, E.F.; Kann, W.J.; Burelbach, J.P.

1984-04-10

Improved thermal insulation for a nuclear reactor deck comprises many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

Status of JUPITER Program

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

Inoue, T.; Shirakata, K.; Kinjo, K.

To obtain the data necessary for evaluating the nuclear design method of a large-scale fast breeder reactor, criticality tests with a large- scale homogeneous reactor were conducted as part of a joint research program by Japan and the U.S. Analyses of the tests are underway in both countries. The purpose of this paper is to describe the status of this project.

NREL's Thermochemical Conversion Facility Video Text Version | Bioenergy |

Science.gov Websites

steady-state. We use a tandem fast pyrolysis reactor and Davison recirculating reactor system to study ex be continually added and withdrawn so we can study catalyst activity and product composition at catalyst. Here we can study the impact of catalyst formulation and processing conditions on bio-oil

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

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

Morreale, A. C.; Ball, M. R.; Novog, D. R.

2012-07-01

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxidemore » fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)« less

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

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

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

10 CFR 110.42 - Export licensing criteria.

Code of Federal Regulations, 2012 CFR

2012-01-01

... research on or development of any nuclear explosive device. (3) Adequate physical security measures will be... to exports of high-enriched uranium to be used as a fuel or target in a nuclear research or test... can be used in the reactor. (iii) A fuel or target “can be used” in a nuclear research or test reactor...

WATER PUMP HOUSE, TRA619, PUMP INSTALLATION. CAMERA FACING NORTHEAST CORNER. ...

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

WATER PUMP HOUSE, TRA-619, PUMP INSTALLATION. CAMERA FACING NORTHEAST CORNER. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON THE ORIGINAL NEGATIVE. INL NEGATIVE NO. 3998. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fast Computation and Assessment Methods in Power System Analysis

NASA Astrophysics Data System (ADS)

Nagata, Masaki

Power system analysis is essential for efficient and reliable power system operation and control. Recently, online security assessment system has become of importance, as more efficient use of power networks is eagerly required. In this article, fast power system analysis techniques such as contingency screening, parallel processing and intelligent systems application are briefly surveyed from the view point of their application to online dynamic security assessment.

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

In-Pile Qualification of the Fast-Neutron-Detection-System

NASA Astrophysics Data System (ADS)

Fourmentel, D.; Villard, J.-F.; Destouches, C.; Geslot, B.; Vermeeren, L.; Schyns, M.

2018-01-01

In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

JAEA's actions and contributions to the strengthening of nuclear non-proliferation

NASA Astrophysics Data System (ADS)

Suda, Kazunori; Suzuki, Mitsutoshi; Michiji, Toshiro

2012-06-01

Japan, a non-nuclear weapons state, has established a commercial nuclear fuel cycle including LWRs, and now is developing a fast neutron reactor fuel cycle as part of the next generation nuclear energy system, with commercial operation targeted for 2050. Japan Atomic Energy Agency (JAEA) is the independent administrative agency for conducting comprehensive nuclear R&D in Japan after the merger of Japan Atomic Energy Research Institute (JAERI) and Japan Nuclear Cycle Development Institute (JNC). JAEA and its predecessors have extensive experience in R&D, facility operations, and safeguards development and implementation for new types of nuclear facilities for the peaceful use of nuclear energy. As the operator of various nuclear fuel cycle facilities and numerous nuclear materials, JAEA makes international contributions to strengthen nuclear non-proliferation. This paper provides an overview of JAEA's development of nuclear non-proliferation and safeguards technologies, including remote monitoring of nuclear facilities, environmental sample analysis methods and new efforts since the 2010 Nuclear Security Summit in Washington D.C.

Mass tracking and material accounting in the Integral Fast Reactor (IFR)

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

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities atmore » ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG Tasks'' which collect, manipulate and report data, (3) a set of MTG Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF.« less

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

[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.

Analysis of the propagation of neutrons and gamma-rays from the fast neutron source reactor YAYOI

NASA Astrophysics Data System (ADS)

Yoshida, Shigeo; Murata, Isao; Nakagawa, Tsutomu; Saito, Isao

2011-10-01

The skyshine effect is crucial for designing appropriate shielding. To investigate the skyshine effect, the propagation of neutrons was measured and analyzed at the fast neutron source reactor YAYOI. Pulse height spectra and dose distributions of neutron and secondary gamma-ray were measured outside YAYOI, and analyzed with MCNP-5 and JENDL-3.3. Comparison with the experimental results showed good agreement. Also, a semi-empirical formula was successfully derived to describe the dose distribution. The formulae can be used to predict the skyshine effect at YAYOI, and will be useful for estimating the skyshine effect and designing the shield structure for fusion facilities.

Impact of nuclear data on sodium-cooled fast reactor calculations

NASA Astrophysics Data System (ADS)

Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

2016-03-01

Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

Bootstrap and fast wave current drive for tokamak reactors

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

Ehst, D.A.

1991-09-01

Using the multi-species neoclassical treatment of Hirshman and Sigmar we study steady state bootstrap equilibria with seed currents provided by low frequency (ICRF) fast waves and with additional surface current density driven by lower hybrid waves. This study applies to reactor plasmas of arbitrary aspect ratio. IN one limit the bootstrap component can supply nearly the total equilibrium current with minimal driving power (< 20 MW). However, for larger total currents considerable driving power is required (for ITER: I{sub o} = 18 MA needs P{sub FW} = 15 MW, P{sub LH} = 75 MW). A computational survey of bootstrap fractionmore » and current drive efficiency is presented. 11 refs., 8 figs.« less

Leveraging Safety Programs to Improve and Support Security Programs

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

Leach, Janice; Snell, Mark K.; Pratt, R.

2015-10-01

There has been a long history of considering Safety, Security, and Safeguards (3S) as three functions of nuclear security design and operations that need to be properly and collectively integrated with operations. This paper specifically considers how safety programmes can be extended directly to benefit security as part of an integrated facility management programme. The discussion will draw on experiences implementing such a programme at Sandia National Laboratories’ Annular Research Reactor Facility. While the paper focuses on nuclear facilities, similar ideas could be used to support security programmes at other types of high-consequence facilities and transportation activities.

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.

Status of French reactors

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

Ballagny, A.

1997-08-01

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

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

NASA Astrophysics Data System (ADS)

Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

2017-09-01

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

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

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated formore » up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.« less

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

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

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

Theoretical Estimate of Maximum Possible Nuclear Explosion

DOE R&D Accomplishments Database

Bethe, H. A.

1950-01-31

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

A CFD model for biomass fast pyrolysis in fluidized-bed reactors

NASA Astrophysics Data System (ADS)

Xue, Qingluan; Heindel, T. J.; Fox, R. O.

2010-11-01

A numerical study is conducted to evaluate the performance and optimal operating conditions of fluidized-bed reactors for fast pyrolysis of biomass to bio-oil. A comprehensive CFD model, coupling a pyrolysis kinetic model with a detailed hydrodynamics model, is developed. A lumped kinetic model is applied to describe the pyrolysis of biomass particles. Variable particle porosity is used to account for the evolution of particle physical properties. The kinetic scheme includes primary decomposition and secondary cracking of tar. Biomass is composed of reference components: cellulose, hemicellulose, and lignin. Products are categorized into groups: gaseous, tar vapor, and solid char. The particle kinetic processes and their interaction with the reactive gas phase are modeled with a multi-fluid model derived from the kinetic theory of granular flow. The gas, sand and biomass constitute three continuum phases coupled by the interphase source terms. The model is applied to investigate the effect of operating conditions on the tar yield in a fluidized-bed reactor. The influence of various parameters on tar yield, including operating temperature and others are investigated. Predicted optimal conditions for tar yield and scale-up of the reactor are discussed.

76 FR 30204 - Exelon Nuclear, Dresden Nuclear Power Station, Unit 1; Exemption From Certain Security Requirements

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-24

... Power Station, Unit 1; Exemption From Certain Security Requirements 1.0 Background Exelon Nuclear is the licensee and holder of Facility Operating License No. DPR-2 issued for Dresden Nuclear Power Station (DNPS... protection of licensed activities in nuclear power reactors against radiological sabotage,'' paragraph (b)(1...

Conceptual design of fast-ignition laser fusion reactor FALCON-D

NASA Astrophysics Data System (ADS)

Goto, T.; Someya, Y.; Ogawa, Y.; Hiwatari, R.; Asaoka, Y.; Okano, K.; Sunahara, A.; Johzaki, T.

2009-07-01

A new conceptual design of the laser fusion power plant FALCON-D (Fast-ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast-ignition method can achieve sufficient fusion gain for a commercial operation (~100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5-6 m radius). 1D/2D simulations by hydrodynamic codes showed a possibility of achieving sufficient gain with a laser energy of 400 kJ, i.e. a 40 MJ target yield. The design feasibility of the compact dry wall chamber and the solid breeder blanket system was shown through thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. Moderate electric output (~400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall reactor concept not only reduces several difficulties associated with a liquid wall system but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance period. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R&D issues required for this design are also discussed.

Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

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

Perko, Z.; Gilli, L.; Lathouwers, D.

2013-07-01

Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used techniquemore » proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)« less

Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel

NASA Astrophysics Data System (ADS)

Edelmann, Paul G.

There have been numerous attempts to model fast reactor fuel behavior in the last 40 years. The US currently does not have a fully reliable tool to simulate the behavior of metal fuels in fast reactors. The experimental database necessary to validate the codes is also very limited. The DOE-sponsored Advanced Fuels Campaign (AFC) has performed various experiments that are ready for analysis. Current metal fuel performance codes are either not available to the AFC or have limitations and deficiencies in predicting AFC fuel performance. A modified version of a new fuel performance code, FEAST-Metal , was employed in this investigation with useful results. This work explores the modeling and analysis of AFC metallic fuels using FEAST-Metal, particularly in the area of constituent actinide diffusion behavior. The FEAST-Metal code calculations for this work were conducted at Los Alamos National Laboratory (LANL) in support of on-going activities related to sensitivity analysis of fuel performance codes. A sensitivity analysis of FEAST-Metal was completed to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. A modification was made to the FEAST-Metal constituent redistribution model to enable accommodation of newer AFC metal fuel compositions with verified results. Applicability of this modified model for sodium fast reactor metal fuel design is demonstrated.

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

75 FR 17196 - Self-Regulatory Organizations; The Depository Trust Company; Order Granting Approval of a...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-05

... Part of the Fast Automated Transfer Program March 30, 2010. I. Introduction On January 19, 2010, The... made unsponsored ADRs eligible for DTC's Fast Automated Securities Transfer Program (``FAST'').\\4\\ \\3... associated with safekeeping, transfer, shipping and insurance costs. \\4\\ FAST was designed to eliminate some...

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

Gehin, Jess C.; Powers, Jeffrey J.

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

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

NASA Astrophysics Data System (ADS)

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

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

Modifications to the NRAD Reactor, 1977 to present

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

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.« less

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

NASA Astrophysics Data System (ADS)

Bailey, Nathan A.; Stergar, Erich; Toloczko, Mychailo; Hosemann, Peter

2015-04-01

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412 °C-109 dpa, chromium enrichments - consistent with the α‧ phase - appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412 °C-109 dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

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

Ponciroli, R.; Passerini, S.; Vilim, R. B.

In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based onmore » the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.« less

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

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

Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa

2002-07-01

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

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

Fast Pyrolysis of Poplar Using a Captive Sample Reactor: Effects of Inorganic Salts on Primary Pyrolysis Products

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

Mukarakate, C.; Robichaud, D.; Donohoe, B.

2012-01-01

We have constructed a captive sample reactor (CSR) to study fast pyrolysis of biomass. The reactor uses a stainless steel wire mesh to surround biomass materials with an isothermal environment by independent controlling of heating rates and pyrolysis temperatures. The vapors produced during pyrolysis are immediately entrained and transported in He carrier gas to a molecular beam mass spectrometer (MBMS). Formation of secondary products is minimized by rapidly quenching the sample support with liquid nitrogen. A range of alkali and alkaline earth metal (AAEM) and transition metal salts were tested to study their effect on composition of primary pyrolysis products.more » Multivariate curve resolution (MCR) analysis of the MBMS data shows that transition metal salts enhance pyrolysis of carbohydrates and AAEM salts enhances pyrolysis of lignin. This was supported by performing similar separate studies on cellulose, hemicellulose and extracted lignin. The effect of salts on char formation is also discussed.« less

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

Cheng, Lap-Yan; Wei, Thomas Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

A novel fast mass transfer anaerobic inner loop fluidized bed biofilm reactor for PTA wastewater treatment.

PubMed

Chen, Yingwen; Zhao, Jinlong; Li, Kai; Xie, Shitao

In this paper, a fast mass transfer anaerobic inner loop fluidized bed biofilm reactor (ILFBBR) was developed to improve purified terephthalic acid (PTA) wastewater treatment. The emphasis of this study was on the start-up mode of the anaerobic ILFBBR, the hydraulic loadings and the operation stability. The biological morphology of the anaerobic biofilm in the reactors was also analyzed. The anaerobic column could operate successfully for 46 days due to the pre-aerating process. The anaerobic column had the capacity to resist shock loadings and maintained a high stable chemical oxygen demand (COD) and terephthalic acid removal rates at a hydraulic retention time of 5-10 h, even under conditions of organic volumetric loadings as high as 28.8 kg COD·m(-3).d(-1). The scanning electron microscope analysis of the anaerobic carrier demonstrated that clusters of prokaryotes grew inside of pores and that the filaments generated by pre-aeration contributed to the anaerobic biofilm formation and stability.

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

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

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

NASA Astrophysics Data System (ADS)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

Performance of low smeared density sodium-cooled fast reactor metal fuel

NASA Astrophysics Data System (ADS)

Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

2015-10-01

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

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

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

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

2013-12-01

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

78 FR 60284 - Submission for OMB Review; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-01

... OMB Review; Comment Request Title: Federally Assisted State Transmitted (FAST) Levy. OMB No.: New... proposing a new information collection using the Federally Assisted State Transmitted Levy (FAST Levy) application. FAST Levy is a centralized, secure and automated method of collecting and disseminating...

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

Mendez Cruz, Carmen Margarita; Rochau, Gary E.; Middleton, Bobby

Sandia National Laboratories and General Atomics are pleased to respond to the Advanced Research Projects Agency-Energy (ARPA-e)’s request for information on innovative developments that may overcome various current reactor-technology limitations. The RFI is particularly interested in innovations that enable ultra-safe and secure modular nuclear energy systems. Our response addresses the specific features for reactor designs called out in the RFI, including a brief assessment of the current state of the technologies that would enable each feature and the methods by which they could be best incorporated into a reactor design.

Thermochemical Process Integration, Scale-Up, and Piloting Publications |

Science.gov Websites

-Economic Assessment of Ex Situ Catalytic Fast Pyrolysis of Biomass: A Fixed Bed Reactor Implementation Scenario for Future Feasibility, Topics in Catalysis Image of a schematic of hot gas filter and ex situ Research Pathways with In Situ and Ex Situ Upgrading of Fast Pyrolysis Vapors, NREL Technical Report Image

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

Benton, J; Wall, D; Parker, E

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

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

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

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

Dual annular rotating "windowed" nuclear reflector reactor control system

DOEpatents

Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

1994-01-01

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

A fast key generation method based on dynamic biometrics to secure wireless body sensor networks for p-health.

PubMed

Zhang, G H; Poon, Carmen C Y; Zhang, Y T

2010-01-01

Body sensor networks (BSNs) have emerged as a new technology for healthcare applications, but the security of communication in BSNs remains a formidable challenge yet to be resolved. The paper discusses the typical attacks faced by BSNs and proposes a fast biometric based approach to generate keys for ensuing confidentiality and authentication in BSN communications. The approach was tested on 900 segments of electrocardiogram. Each segment was 4 seconds long and used to generate a 128-bit key. The results of the study found that entropy of 96% of the keys were above 0.95 and 99% of the hamming distances calculated from any two keys were above 50 bits. Based on the randomness and distinctiveness of these keys, it is concluded that the fast biometric based approach has great potential to be used to secure communication in BSNs for health applications.

High-performance compression and double cryptography based on compressive ghost imaging with the fast Fourier transform

NASA Astrophysics Data System (ADS)

Leihong, Zhang; Zilan, Pan; Luying, Wu; Xiuhua, Ma

2016-11-01

To solve the problem that large images can hardly be retrieved for stringent hardware restrictions and the security level is low, a method based on compressive ghost imaging (CGI) with Fast Fourier Transform (FFT) is proposed, named FFT-CGI. Initially, the information is encrypted by the sender with FFT, and the FFT-coded image is encrypted by the system of CGI with a secret key. Then the receiver decrypts the image with the aid of compressive sensing (CS) and FFT. Simulation results are given to verify the feasibility, security, and compression of the proposed encryption scheme. The experiment suggests the method can improve the quality of large images compared with conventional ghost imaging and achieve the imaging for large-sized images, further the amount of data transmitted largely reduced because of the combination of compressive sensing and FFT, and improve the security level of ghost images through ciphertext-only attack (COA), chosen-plaintext attack (CPA), and noise attack. This technique can be immediately applied to encryption and data storage with the advantages of high security, fast transmission, and high quality of reconstructed information.

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

NASA Astrophysics Data System (ADS)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

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

Jaffke, Patrick John

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

Ya B Zeldovich and nuclear power

NASA Astrophysics Data System (ADS)

Ponomarev, L. I.

2014-03-01

The idea on a homogeneous nuclear reactor, first suggested by Ya B Zeldovich and Yu B Khariton in 1939, has since had its ups and downs and is now re-emerging, enriched with the knowledge and experience accumulated over the years having past. One of the current versions of the idea, the fast molten-salt reactor with a U-Pu fuel cycle, is presented in this paper.

Operational performance of the three bean salad control algorithm on the ACRR

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

Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute.

Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

DOEpatents

Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

1999-01-01

A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

Operational performance of the three bean salad control algorithm on the ACRR

NASA Astrophysics Data System (ADS)

Ball, Russell M.; Madaras, John J.; Trowbridge, F. Ray; Talley, Darren G.; Parma, Edward J.

1991-01-01

Experimental tests on the Annular Core Research Reactor have confirmed that the ``Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute.

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

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

PubMed

Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

2010-10-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

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

Industrial research for transmutation scenarios

NASA Astrophysics Data System (ADS)

Camarcat, Noel; Garzenne, Claude; Le Mer, Joël; Leroyer, Hadrien; Desroches, Estelle; Delbecq, Jean-Michel

2011-04-01

This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDF's suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

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

Paumel, Kevin; Lhuillier, Christian

2015-07-01

Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depthmore » seems minor in the range under investigation. (authors)« less

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Progress on inert matrix fuels for minor actinide transmutation in fast reactor

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

Bonnerot, Jean-Marc; Ferroud-Plattet, Marie-Pierre; Lamontagne, Jerome

2007-07-01

An extensive irradiation program has been devoted by CEA to the assessment of transmutation using minor actinide bearing inert support targets. A first irradiation experiment was performed in the fast neutron reactor Phenix, in parallel to other experiments carried out in the HFR and Siloe reactors, in order to assess the behavior under fast neutron flux of various materials intended as inert support matrix for transmutation targets. This experiment, which included the two steps MATINA 1 and MATINA 1A, was completed in 2004 and underwent complete post irradiation examinations (PIE) , whose results are presented in this paper. All themore » pure inert materials showed a satisfactory behavior under fast neutrons except Al{sub 2}O{sub 3} - which exhibits a swelling close to 11 vol. % after irradiation. In presence of UO{sub 2} fissile particles, MgAl{sub 2}O{sub 4} proved to be more stable in term of swelling as inert support than MgO and Al{sub 2}O{sub 3} matrices, under the same irradiation conditions. A second experiment ECRIX H in Phenix involving composite pellets with an MgO matrix and AmO{sub 2-x} particles was completed in 2006. The very first PIE results on ECRIX H are described in this paper. At the light of these first experiments, a second phase dedicated to the design optimization of the target was initiated and three new irradiation experiments - MATINA 2-3, CAMIX COCHIX in Phenix and HELIOS in HFR - were started in 2006 and 2007. (authors)« less

Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

NASA Astrophysics Data System (ADS)

Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

The necessity of nuclear reactors for targeted radionuclide therapies.

PubMed

Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

2013-07-01

Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place. Copyright © 2013 Elsevier Ltd. All rights reserved.

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

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

Lee, Changho; Yang, Won Sik

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400more » 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.« less

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

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

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

DEMINERALIZER BUILDING, TRA608. INSTALLATION OF SAMPLING AND OTHER INSTRUMENTS COMPLETES ...

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

DEMINERALIZER BUILDING, TRA-608. INSTALLATION OF SAMPLING AND OTHER INSTRUMENTS COMPLETES DEMINERALIZER UNITS ALONG NORTH WALL. CAMERA FACES EAST. CARD IN LOWER RIGHT WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON THE ORIGINAL NEGATIVE. INL NEGATIVE NO. 3996A. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING, TRA604. A LABORATORY ROOM WITH ITS CABINETS AND ...

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

MTR WING, TRA-604. A LABORATORY ROOM WITH ITS CABINETS AND SERVICE STRIP DOWN CENTER OF ROOM. CARD IN LEFT CORNER OF VIEW WAS INSERTED BY INL PHOTOGRAPHER TO COVER AN OBSOLETE SECURITY RESTRICTION PRINTED ON THE ORIGINAL NEGATIVE. INL NEGATIVE NO. 3817. Unknown Photographer, 11/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

2016-05-01

The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.