ITER activities and fusion technology

NASA Astrophysics Data System (ADS)

Seki, M.

2007-10-01

At the 21st IAEA Fusion Energy Conference, 68 and 67 papers were presented in the categories of ITER activities and fusion technology, respectively. ITER performance prediction, results of technology R&D and the construction preparation provide good confidence in ITER realization. The superconducting tokamak EAST achieved the first plasma just before the conference. The construction of other new experimental machines has also shown steady progress. Future reactor studies stress the importance of down sizing and a steady-state approach. Reactor technology in the field of blanket including the ITER TBM programme and materials for the demonstration power plant showed sound progress in both R&D and design activities.

Overview of the present progress and activities on the CFETR

NASA Astrophysics Data System (ADS)

Wan, Yuanxi; Li, Jiangang; Liu, Yong; Wang, Xiaolin; Chan, Vincent; Chen, Changan; Duan, Xuru; Fu, Peng; Gao, Xiang; Feng, Kaiming; Liu, Songlin; Song, Yuntao; Weng, Peide; Wan, Baonian; Wan, Farong; Wang, Heyi; Wu, Songtao; Ye, Minyou; Yang, Qingwei; Zheng, Guoyao; Zhuang, Ge; Li, Qiang; CFETR Team

2017-10-01

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

Status and improvement of CLAM for nuclear application

NASA Astrophysics Data System (ADS)

Huang, Qunying

2017-08-01

A program for China low activation martensitic steel (CLAM) development has been underway since 2001 to satisfy the material requirements of the test blanket module (TBM) for ITER, China fusion engineering test reactor and China fusion demonstration reactor. It has been undertaken by the Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences under wide domestic and international collaborations. Extensive work and efforts are being devoted to the R&D of CLAM, such as mechanical property evaluation before and after neutron irradiation, fabrication of scaled TBM by welding and additive manufacturing, improvement of its irradiation resistance as well as high temperature properties by precipitate strengthening to achieve its final successful application in fusion systems. The status and improvement of CLAM are introduced in this paper.

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

NASA Astrophysics Data System (ADS)

H, L. SWAMI; C, DANANI; A, K. SHAW

2018-06-01

Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a high-energy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic‑Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well as the dose level, which has an impact on design considerations. IN-RAFMS was shown to be a more effective low-activation material than SS-316LN-IG.

Armour Materials for the ITER Plasma Facing Components

NASA Astrophysics Data System (ADS)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

NASA Astrophysics Data System (ADS)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

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

Bourham, Mohamed A.; Gilligan, John G.

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing componentsmore » safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.« less

Eigenvalue Solvers for Modeling Nuclear Reactors on Leadership Class Machines

DOE PAGES

Slaybaugh, R. N.; Ramirez-Zweiger, M.; Pandya, Tara; ...

2018-02-20

In this paper, three complementary methods have been implemented in the code Denovo that accelerate neutral particle transport calculations with methods that use leadership-class computers fully and effectively: a multigroup block (MG) Krylov solver, a Rayleigh quotient iteration (RQI) eigenvalue solver, and a multigrid in energy (MGE) preconditioner. The MG Krylov solver converges more quickly than Gauss Seidel and enables energy decomposition such that Denovo can scale to hundreds of thousands of cores. RQI should converge in fewer iterations than power iteration (PI) for large and challenging problems. RQI creates shifted systems that would not be tractable without the MGmore » Krylov solver. It also creates ill-conditioned matrices. The MGE preconditioner reduces iteration count significantly when used with RQI and takes advantage of the new energy decomposition such that it can scale efficiently. Each individual method has been described before, but this is the first time they have been demonstrated to work together effectively. The combination of solvers enables the RQI eigenvalue solver to work better than the other available solvers for large reactors problems on leadership-class machines. Using these methods together, RQI converged in fewer iterations and in less time than PI for a full pressurized water reactor core. These solvers also performed better than an Arnoldi eigenvalue solver for a reactor benchmark problem when energy decomposition is needed. The MG Krylov, MGE preconditioner, and RQI solver combination also scales well in energy. Finally, this solver set is a strong choice for very large and challenging problems.« less

Eigenvalue Solvers for Modeling Nuclear Reactors on Leadership Class Machines

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

Slaybaugh, R. N.; Ramirez-Zweiger, M.; Pandya, Tara

In this paper, three complementary methods have been implemented in the code Denovo that accelerate neutral particle transport calculations with methods that use leadership-class computers fully and effectively: a multigroup block (MG) Krylov solver, a Rayleigh quotient iteration (RQI) eigenvalue solver, and a multigrid in energy (MGE) preconditioner. The MG Krylov solver converges more quickly than Gauss Seidel and enables energy decomposition such that Denovo can scale to hundreds of thousands of cores. RQI should converge in fewer iterations than power iteration (PI) for large and challenging problems. RQI creates shifted systems that would not be tractable without the MGmore » Krylov solver. It also creates ill-conditioned matrices. The MGE preconditioner reduces iteration count significantly when used with RQI and takes advantage of the new energy decomposition such that it can scale efficiently. Each individual method has been described before, but this is the first time they have been demonstrated to work together effectively. The combination of solvers enables the RQI eigenvalue solver to work better than the other available solvers for large reactors problems on leadership-class machines. Using these methods together, RQI converged in fewer iterations and in less time than PI for a full pressurized water reactor core. These solvers also performed better than an Arnoldi eigenvalue solver for a reactor benchmark problem when energy decomposition is needed. The MG Krylov, MGE preconditioner, and RQI solver combination also scales well in energy. Finally, this solver set is a strong choice for very large and challenging problems.« less

Fusion materials: Technical evaluation of the technology of vandium alloys for use as blanket structural materials in fusion power systems

NASA Astrophysics Data System (ADS)

1993-08-01

The Committee's evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is limited and will require an extensive development and test program. Because of the chemical reactivity of vanadium the alloy will be less tolerant of system failures, accidents, and off-normal events than most other candidate blanket structural materials and will require more careful handling during fabrication of hardware. Because of the cost of the material more stringent requirements on processes, and minimal historical working experience, it will cost an order of magnitude to qualify a vanadium alloy for ITER blanket structures than other candidate materials. The use of vanadium is difficult and uncertain; therefore, other options should be explored more thoroughly before a final selection of vanadium is confirmed. The Committee views the risk as being too high to rely solely on vanadium alloys. In viewing the state and nature of the design of the ITER blanket as presented to the Committee, it is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium alloy option is to be pursued, a large program needs to be started immediately. The commitment of funding and other resources needs to be firm and consistent with a realistic program plan.

Radioactivity measurements of ITER materials using the TFTR D-T neutron field

NASA Astrophysics Data System (ADS)

Kumar, A.; Abdou, M. A.; Barnes, C. W.; Kugel, H. W.

1994-06-01

The availability of high D-T fusion neutron yields at TFTR has provided a useful opportunity to directly measure D-T neutron-induced radioactivity in a realistic tokamak fusion reactor environment for materials of vital interest to ITER. These measurements are valuable for characterizing radioactivity in various ITER candidate materials, for validating complex neutron transport calculations, and for meeting fusion reactor licensing requirements. The radioactivity measurements at TFTR involve potential ITER materials including stainless steel 316, vanadium, titanium, chromium, silicon, iron, cobalt, nickel, molybdenum, aluminum, copper, zinc, zirconium, niobium, and tungsten. Small samples of these materials were irradiated close to the plasma and just outside the vacuum vessel wall of TFTR, locations of different neutron energy spectra. Saturation activities for both threshold and capture reactions were measured. Data from dosimetric reactions have been used to obtain preliminary neutron energy spectra. Spectra from the first wall were compared to calculations from ITER and to measurements from accelerator-based tests.

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

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

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Constitutive law for thermally-activated plasticity of recrystallized tungsten

NASA Astrophysics Data System (ADS)

Zinovev, Aleksandr; Terentyev, Dmitry; Dubinko, Andrii; Delannay, Laurent

2017-12-01

A physically-based constitutive law relevant for ITER-specification tungsten grade in as-recrystallized state is proposed. The material demonstrates stages III and IV of the plastic deformation, in which hardening rate does not drop to zero with the increase of applied stress. Despite the classical Kocks-Mecking model, valid at stage III, the strain hardening asymptotically decreases resembling a hyperbolic function. The material parameters are fitted by relying on tensile test data and by requiring that the strain and stress at the onset of diffuse necking (uniform elongation and ultimate tensile strength correspondingly) as well as the yield stress be reproduced. The model is then validated in the temperature range 300-600 °C with the help of finite element analysis of tensile tests which confirms the reproducibility of the experimental engineering curves up to the onset of diffuse necking, beyond which the development of ductile damage accelerates the material failure. This temperature range represents the low temperature application window for tungsten as divertor material in fusion reactor ITER.

Two-dimensional over-all neutronics analysis of the ITER device

NASA Astrophysics Data System (ADS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

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.

Tritium proof-of-principle pellet injector: Phase 2

NASA Astrophysics Data System (ADS)

Fisher, P. W.; Gouge, M. J.

1995-03-01

As part of the International Thermonuclear Engineering Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. This repeating, single-stage, pneumatic injector, called the Tritium-Proof-of-Principle Phase-2 (TPOP-2) Pellet Injector, has a piston-driven mechanical extruder and is designed to extrude hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and DT mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and DT extrusions; integrate, test and evaluate the extruder in a repeating, single-stage light gas gun sized for the ITER application (pellet diameter approximately 7-8 mm); evaluate options for recycling propellant and extruder exhaust gas; evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory requiring secondary and room containment systems. In initial tests with deuterium feed at ORNL, up to thirteen pellets have been extruded at rates up to 1 Hz and accelerated to speeds of order 1.0-1.1 km/s using hydrogen propellant gas at a supply pressure of 65 bar. The pellets are typically 7.4 mm in diameter and up to 11 mm in length and are the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 11% density perturbation to ITER. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first wall tritium inventories by a process called isotopic fueling where tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Fusion Power measurement at ITER

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

Bertalot, L.; Barnsley, R.; Krasilnikov, V.

2015-07-01

Nuclear fusion research aims to provide energy for the future in a sustainable way and the ITER project scope is to demonstrate the feasibility of nuclear fusion energy. ITER is a nuclear experimental reactor based on a large scale fusion plasma (tokamak type) device generating Deuterium - Tritium (DT) fusion reactions with emission of 14 MeV neutrons producing up to 700 MW fusion power. The measurement of fusion power, i.e. total neutron emissivity, will play an important role for achieving ITER goals, in particular the fusion gain factor Q related to the reactor performance. Particular attention is given also tomore » the development of the neutron calibration strategy whose main scope is to achieve the required accuracy of 10% for the measurement of fusion power. Neutron Flux Monitors located in diagnostic ports and inside the vacuum vessel will measure ITER total neutron emissivity, expected to range from 1014 n/s in Deuterium - Deuterium (DD) plasmas up to almost 10{sup 21} n/s in DT plasmas. The neutron detection systems as well all other ITER diagnostics have to withstand high nuclear radiation and electromagnetic fields as well ultrahigh vacuum and thermal loads. (authors)« less

Fusion energy

NASA Astrophysics Data System (ADS)

1990-09-01

The main purpose of the International Thermonuclear Experimental Reactor (ITER) is to develop an experimental fusion reactor through the united efforts of many technologically advanced countries. The ITER terms of reference, issued jointly by the European Community, Japan, the USSR, and the United States, call for an integrated international design activity and constitute the basis of current activities. Joint work on ITER is carried out under the auspices of the International Atomic Energy Agency (IAEA), according to the terms of quadripartite agreement reached between the European Community, Japan, the USSR, and the United States. The site for joint technical work sessions is at the Max Planck Institute of Plasma Physics. Garching, Federal Republic of Germany. The ITER activities have two phases: a definition phase performed in 1988 and the present design phase (1989 to 1990). During the definition phase, a set of ITER technical characteristics and supporting research and development (R and D) activities were developed and reported. The present conceptual design phase of ITER lasts until the end of 1990. The objectives of this phase are to develop the design of ITER, perform a safety and environmental analysis, develop site requirements, define future R and D needs, and estimate cost, manpower, and schedule for construction and operation. A final report will be submitted at the end of 1990. This paper summarizes progress in the ITER program during the 1989 design phase.

A Burning Plasma Experiment: the role of international collaboration

NASA Astrophysics Data System (ADS)

Prager, Stewart

2003-04-01

The world effort to develop fusion energy is at the threshold of a new stage in its research: the investigation of burning plasmas. A burning plasma is self-heated. The 100 million degree temperature of the plasma is maintained by the heat generated by the fusion reactions themselves, as occurs in burning stars. The fusion-generated alpha particles produce new physical phenomena that are strongly coupled together as a nonlinear complex system, posing a major plasma physics challenge. Two attractive options are being considered by the US fusion community as burning plasma facilities: the international ITER experiment and the US-based FIRE experiment. ITER (the International Thermonuclear Experimental Reactor) is a large, power-plant scale facility. It was conceived and designed by a partnership of the European Union, Japan, the Soviet Union, and the United States. At the completion of the first engineering design in 1998, the US discontinued its participation. FIRE (the Fusion Ignition Research Experiment) is a smaller, domestic facility that is at an advanced pre-conceptual design stage. Each facility has different scientific, programmatic and political implications. Selecting the optimal path for burning plasma science is itself a challenge. Recently, the Fusion Energy Sciences Advisory Committee recommended a dual path strategy in which the US seek to rejoin ITER, but be prepared to move forward with FIRE if the ITER negotiations do not reach fruition by July, 2004. Either the ITER or FIRE experiment would reveal the behavior of burning plasmas, generate large amounts of fusion power, and be a huge step in establishing the potential of fusion energy to contribute to the world's energy security.

Fusion Breeding for Sustainable, Mid Century, Carbon Free Power

NASA Astrophysics Data System (ADS)

Manheimer, Wallace

2015-11-01

If ITER achieves Q ~10, it is still very far from useful fusion. The fusion power, and the driver power will allow only a small amount of power to be delivered, <~50MW for an ITER scale tokamak. It is unlikely, considering ``conservative design rules'' that tokamaks can ever be economical pure fusion power producers. Considering the status of other magnetic fusion concepts, it is also very unlikely that any alternate concept will either. Laser fusion does not seem to be constrained by any conservative design rules, but considering the failure of NIF to achhieve ignition, at this point it has many more obstacles to overcome than magnetic fusion. One way out of this dilemma is to use an ITER size tokamak, or a NIF size laser, as a fuel breeder for searate nuclear reactors. Hence ITER and NIF become ends in themselves, instead of steps to who knows what DEMO decades later. Such a tokamak can easily live within the consrtaints of conservative design rules. This has led the author to propose ``The Energy Park'' a sustainable, carbon free, economical, and environmently viable power source without prolifertion risk. It is one fusion breeder fuels 5 conventional nuclear reactors, and one fast neutron reactor burns the actinide wastes.

Conceptual design of the cryogenic system and estimation of the recirculated power for CFETR

NASA Astrophysics Data System (ADS)

Liu, Xiaogang; Qiu, Lilong; Li, Junjun; Wang, Zhaoliang; Ren, Yong; Wang, Xianwei; Li, Guoqiang; Gao, Xiang; Bi, Yanfang

2017-01-01

The China Fusion Engineering Test Reactor (CFETR) is the next tokamak in China’s roadmap for realizing commercial fusion energy. The CFETR cryogenic system is crucial to creating and maintaining operational conditions for its superconducting magnet system and thermal shields. The preliminary conceptual design of the CFETR cryogenic system has been carried out with reference to that of ITER. It will provide an average capacity of 75 to 80 kW at 4.5 K and a peak capacity of 1300 kW at 80 K. The electric power consumption of the cryogenic system is estimated to be 24 MW, and the gross building area is about 7000 m2. The relationships among the auxiliary power consumed by the cryogenic system, the fusion power gain and the recirculated power of CFETR are discussed, with the suggestion that about 52% of the electric power produced by CFETR in phase II must be recirculated to run the fusion test reactor.

Development of High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Farmer, J.; Dixon, D.; Kapernick, R.; Dickens, R.; Adams, M.

2007-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Work at the NASA Marshall Space Flight Center seeks to develop high fidelity thermal simulators that not only match the static power profile that would be observed in an operating, fueled nuclear reactor, but to also match the dynamic fuel pin performance during feasible transients. Comparison between the fuel pins and thermal simulators is made at the fuel clad surface, which corresponds to the sheath surface in the thermal simulator. Static and dynamic fuel pin performance was determined using SINDA-FLUINT analysis, and the performance of conceptual thermal simulator designs was compared to the expected nuclear performance. Through a series of iterative analysis, a conceptual high fidelity design will be developed, followed by engineering design, fabrication, and testing to validate the overall design process. Although the resulting thermal simulator will be designed for a specific reactor concept, establishing this rigorous design process will assist in streamlining the thermal simulator development for other reactor concepts.

The engineering design of the Tokamak Physics Experiment

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

Schmidt, J.A.

A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design.more » If adequately funded the construction project should be completed in the year 2000.« less

Parallel computation of multigroup reactivity coefficient using iterative method

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

Iteration in Early-Elementary Engineering Design

ERIC Educational Resources Information Center

McFarland Kendall, Amber Leigh

2017-01-01

K-12 standards and curricula are beginning to include engineering design as a key practice within Science Technology Engineering and Mathematics (STEM) education. However, there is little research on how the youngest students engage in engineering design within the elementary classroom. This dissertation focuses on iteration as an essential aspect…

Iteration in Early-Elementary Engineering Design

NASA Astrophysics Data System (ADS)

McFarland Kendall, Amber Leigh

K-12 standards and curricula are beginning to include engineering design as a key practice within Science Technology Engineering and Mathematics (STEM) education. However, there is little research on how the youngest students engage in engineering design within the elementary classroom. This dissertation focuses on iteration as an essential aspect of engineering design, and because research at the college and professional level suggests iteration improves the designer's understanding of problems and the quality of design solutions. My research presents qualitative case studies of students in kindergarten and third-grade as they engage in classroom engineering design challenges which integrate with traditional curricula standards in mathematics, science, and literature. I discuss my results through the lens of activity theory, emphasizing practices, goals, and mediating resources. Through three chapters, I provide insight into how early-elementary students iterate upon their designs by characterizing the ways in which lesson design impacts testing and revision, by analyzing the plan-driven and experimentation-driven approaches that student groups use when solving engineering design challenges, and by investigating how students attend to constraints within the challenge. I connect these findings to teacher practices and curriculum design in order to suggest methods of promoting iteration within open-ended, classroom-based engineering design challenges. This dissertation contributes to the field of engineering education by providing evidence of productive engineering practices in young students and support for the value of engineering design challenges in developing students' participation and agency in these practices.

Note: Readout of a micromechanical magnetometer for the ITER fusion reactor.

PubMed

Rimminen, H; Kyynäräinen, J

2013-05-01

We present readout instrumentation for a MEMS magnetometer, placed 30 m away from the MEMS element. This is particularly useful when sensing is performed in high-radiation environment, where the semiconductors in the readout cannot survive. High bandwidth transimpedance amplifiers are used to cancel the cable capacitances of several nanofarads. A frequency doubling readout scheme is used for crosstalk elimination. Signal-to-noise ratio in the range of 60 dB was achieved and with sub-percent nonlinearity. The presented instrument is intended for the steady-state magnetic field measurements in the ITER fusion reactor.

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

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

Kelley, D.B.

1995-12-31

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

Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Dufek, Jan

2014-06-01

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

Heat transfer evaluation in a plasma core reactor

NASA Technical Reports Server (NTRS)

Smith, D. E.; Smith, T. M.; Stoenescu, M. L.

1976-01-01

Numerical evaluations of heat transfer in a fissioning uranium plasma core reactor cavity, operating with seeded hydrogen propellant, was performed. A two-dimensional analysis is based on an assumed flow pattern and cavity wall heat exchange rate. Various iterative schemes were required by the nature of the radiative field and by the solid seed vaporization. Approximate formulations of the radiative heat flux are generally used, due to the complexity of the solution of a rigorously formulated problem. The present work analyzes the sensitivity of the results with respect to approximations of the radiative field, geometry, seed vaporization coefficients and flow pattern. The results present temperature, heat flux, density and optical depth distributions in the reactor cavity, acceptable simplifying assumptions, and iterative schemes. The present calculations, performed in cartesian and spherical coordinates, are applicable to any most general heat transfer problem.

Final Report on ITER Task Agreement 81-10

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

Brad J. Merrill

An International Thermonuclear Experimental Reactor (ITER) Implementing Task Agreement (ITA) on Magnet Safety was established between the ITER International Organization (IO) and the Idaho National Laboratory (INL) Fusion Safety Program (FSP) during calendar year 2004. The objectives of this ITA were to add new capabilities to the MAGARC code and to use this updated version of MAGARC to analyze unmitigated superconductor quench events for both poloidal field (PF) and toroidal field (TF) coils of the ITER design. This report documents the completion of the work scope for this ITA. Based on the results obtained for this ITA, an unmitigated quenchmore » event in an ITER larger PF coil does not appear to be as severe an accident as in an ITER TF coil.« less

Design of the helium cooled lithium lead breeding blanket in CEA: from TBM to DEMO

NASA Astrophysics Data System (ADS)

Aiello, G.; Aubert, J.; Forest, L.; Jaboulay, J.-C.; Li Puma, A.; Boccaccini, L. V.

2017-04-01

The helium cooled lithium lead (HCLL) blanket concept was originally developed in CEA at the beginning of 2000: it is one of the two European blanket concepts to be tested in ITER in the form of a test blanket module (TBM) and one of the four blanket concepts currently being considered for the DEMOnstration reactor that will follow ITER. The TBM is a highly optimized component for the ITER environment that will provide crucial information for the development of the DEMO blanket, but its design needs to be adapted to the DEMO reactor. With respect to the TBM design, reduction of the steel content in the breeding zone (BZ) is sought in order to maximize tritium breeding reactions. Different options are being studied, with the potential of reaching tritium breeding ratio (TBR) values up to 1.21. At the same time, the design of the back supporting structure (BSS), which is a DEMO specific component that has to support the blanket modules inside the vacuum vessel (VV), is ongoing with the aim of maximizing the shielding power and minimizing pumping power. This implies a re-engineering of the modules’ attachment system. Design changes however, will have an impact on the manufacturing and assembly sequences that are being developed for the HCLL-TBM. Due to the differences in joint configurations, thicknesses to be welded, heat dissipation and the various technical constraints related to the accessibility of the welding tools and implementation of non-destructive examination (NDE), the manufacturing procedure should be adapted and optimized for DEMO design. Laser welding instead of TIG could be an option to reduce distortions. The time-of-flight diffraction (TOFD) technique is being investigated for NDE. Finally, essential information expected from the HCLL-TBM program that will be needed to finalize the DEMO design is discussed.

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

EPA Science Inventory

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

Developing DIII-D To Prepare For ITER And The Path To Fusion Energy

NASA Astrophysics Data System (ADS)

Buttery, Richard; Hill, David; Solomon, Wayne; Guo, Houyang; DIII-D Team

2017-10-01

DIII-D pursues the advancement of fusion energy through scientific understanding and discovery of solutions. Research targets two key goals. First, to prepare for ITER we must resolve how to use its flexible control tools to rapidly reach Q =10, and develop the scientific basis to interpret results from ITER for fusion projection. Second, we must determine how to sustain a high performance fusion core in steady state conditions, with minimal actuators and a plasma exhaust solution. DIII-D will target these missions with: (i) increased electron heating and balanced torque neutral beams to simulate burning plasma conditions (ii) new 3D coil arrays to resolve control of transients (iii) off axis current drive to study physics in steady state regimes (iv) divertors configurations to promote detachment with low upstream density (v) a reactor relevant wall to qualify materials and resolve physics in reactor-like conditions. With new diagnostics and leading edge simulation, this will position the US for success in ITER and a unique knowledge to accelerate the approach to fusion energy. Supported by the US DOE under DE-FC02-04ER54698.

Rigorous-two-Steps scheme of TRIPOLI-4® Monte Carlo code validation for shutdown dose rate calculation

NASA Astrophysics Data System (ADS)

Jaboulay, Jean-Charles; Brun, Emeric; Hugot, François-Xavier; Huynh, Tan-Dat; Malouch, Fadhel; Mancusi, Davide; Tsilanizara, Aime

2017-09-01

After fission or fusion reactor shutdown the activated structure emits decay photons. For maintenance operations the radiation dose map must be established in the reactor building. Several calculation schemes have been developed to calculate the shutdown dose rate. These schemes are widely developed in fusion application and more precisely for the ITER tokamak. This paper presents the rigorous-two-steps scheme implemented at CEA. It is based on the TRIPOLI-4® Monte Carlo code and the inventory code MENDEL. The ITER shutdown dose rate benchmark has been carried out, results are in a good agreement with the other participant.

The Effect of Iteration on the Design Performance of Primary School Children

ERIC Educational Resources Information Center

Looijenga, Annemarie; Klapwijk, Remke; de Vries, Marc J.

2015-01-01

Iteration during the design process is an essential element. Engineers optimize their design by iteration. Research on iteration in Primary Design Education is however scarce; possibly teachers believe they do not have enough time for iteration in daily classroom practices. Spontaneous playing behavior of children indicates that iteration fits in…

ITER Magnet Feeder: Design, Manufacturing and Integration

NASA Astrophysics Data System (ADS)

CHEN, Yonghua; ILIN, Y.; M., SU; C., NICHOLAS; BAUER, P.; JAROMIR, F.; LU, Kun; CHENG, Yong; SONG, Yuntao; LIU, Chen; HUANG, Xiongyi; ZHOU, Tingzhi; SHEN, Guang; WANG, Zhongwei; FENG, Hansheng; SHEN, Junsong

2015-03-01

The International Thermonuclear Experimental Reactor (ITER) feeder procurement is now well underway. The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in the last 2 years along with analyses and qualification activities. The feeder design is being progressively finalized. In addition, the preparation of qualification and manufacturing are well scheduled at ASIPP. This paper mainly presents the design, the overview of manufacturing and the status of integration on the ITER magnet feeders. supported by the National Special Support for R&D on Science and Technology for ITER (Ministry of Public Security of the People's Republic of China-MPS) (No. 2008GB102000)

The Physics Basis of ITER Confinement

NASA Astrophysics Data System (ADS)

Wagner, F.

2009-02-01

ITER will be the first fusion reactor and the 50 year old dream of fusion scientists will become reality. The quality of magnetic confinement will decide about the success of ITER, directly in the form of the confinement time and indirectly because it decides about the plasma parameters and the fluxes, which cross the separatrix and have to be handled externally by technical means. This lecture portrays some of the basic principles which govern plasma confinement, uses dimensionless scaling to set the limits for the predictions for ITER, an approach which also shows the limitations of the predictions, and describes briefly the major characteristics and physics behind the H-mode—the preferred confinement regime of ITER.

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

Internal combustion engine having a reactor for afterburning of unburned exhaust gas constituents

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

Maurhoff, G.; Steinwart, J.

1974-08-07

An internal combustion engine is described which has an engine housing and a reactor for afterburning of unburned constituents in the exhaust gas. The reactor has a shell with a periphery and contains a heat-insulated, reactor chamber which is freely movable beyond the point of connection to the shell. The reactor has an inlet nozzle extending freely through the shell and connected to an outlet passage of the engine and has an outlet for escape of the exhaust gases from the reactor chamber. The inlet nozzle protrudes freely into the outlet passage, and the shell has a portion around themore » inlet nozzle in contact with the engine housing.« less

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

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

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

Remote experimental site concept development

NASA Astrophysics Data System (ADS)

Casper, Thomas A.; Meyer, William; Butner, David

1995-01-01

Scientific research is now often conducted on large and expensive experiments that utilize collaborative efforts on a national or international scale to explore physics and engineering issues. This is particularly true for the current US magnetic fusion energy program where collaboration on existing facilities has increased in importance and will form the basis for future efforts. As fusion energy research approaches reactor conditions, the trend is towards fewer large and expensive experimental facilities, leaving many major institutions without local experiments. Since the expertise of various groups is a valuable resource, it is important to integrate these teams into an overall scientific program. To sustain continued involvement in experiments, scientists are now often required to travel frequently, or to move their families, to the new large facilities. This problem is common to many other different fields of scientific research. The next-generation tokamaks, such as the Tokamak Physics Experiment (TPX) or the International Thermonuclear Experimental Reactor (ITER), will operate in steady-state or long pulse mode and produce fluxes of fusion reaction products sufficient to activate the surrounding structures. As a direct consequence, remote operation requiring robotics and video monitoring will become necessary, with only brief and limited access to the vessel area allowed. Even the on-site control room, data acquisition facilities, and work areas will be remotely located from the experiment, isolated by large biological barriers, and connected with fiber-optics. Current planning for the ITER experiment includes a network of control room facilities to be located in the countries of the four major international partners; USA, Russian Federation, Japan, and the European Community.

A New Capability for Nuclear Thermal Propulsion Design

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

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

ENGINEERING TEST REACTOR (ETR) BUILDING, TRA642. CONTEXTUAL VIEW, CAMERA FACING ...

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

ENGINEERING TEST REACTOR (ETR) BUILDING, TRA-642. CONTEXTUAL VIEW, CAMERA FACING EAST. VERTICAL METAL SIDING. ROOF IS SLIGHTLY ELEVATED AT CENTER LINE FOR DRAINAGE. WEST SIDE OF ETR COMPRESSOR BUILDING, TRA-643, PROJECTS TOWARD LEFT AT FAR END OF ETR BUILDING. INL NEGATIVE NO. HD46-37-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Efficient solution of the simplified P N equations

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.

2014-12-23

We show new solver strategies for the multigroup SPN equations for nuclear reactor analysis. By forming the complete matrix over space, moments, and energy a robust set of solution strategies may be applied. Moreover, power iteration, shifted power iteration, Rayleigh quotient iteration, Arnoldi's method, and a generalized Davidson method, each using algebraic and physics-based multigrid preconditioners, have been compared on C5G7 MOX test problem as well as an operational PWR model. These results show that the most ecient approach is the generalized Davidson method, that is 30-40 times faster than traditional power iteration and 6-10 times faster than Arnoldi's method.

Conceptual design of the DEMO neutral beam injectors: main developments and R&D achievements

NASA Astrophysics Data System (ADS)

Sonato, P.; Agostinetti, P.; Bolzonella, T.; Cismondi, F.; Fantz, U.; Fassina, A.; Franke, T.; Furno, I.; Hopf, C.; Jenkins, I.; Sartori, E.; Tran, M. Q.; Varje, J.; Vincenzi, P.; Zanotto, L.

2017-05-01

The objectives of the nuclear fusion power plant DEMO, to be built after the ITER experimental reactor, are usually understood to lie somewhere between those of ITER and a ‘first of a kind’ commercial plant. Hence, in DEMO the issues related to efficiency and RAMI (reliability, availability, maintainability and inspectability) are among the most important drivers for the design, as the cost of the electricity produced by this power plant will strongly depend on these aspects. In the framework of the EUROfusion Work Package Heating and Current Drive within the Power Plant Physics and Development activities, a conceptual design of the neutral beam injector (NBI) for the DEMO fusion reactor has been developed by Consorzio RFX in collaboration with other European research institutes. In order to improve efficiency and RAMI aspects, several innovative solutions have been introduced in comparison to the ITER NBI, mainly regarding the beam source, neutralizer and vacuum pumping systems.

Characterization of the ITER model negative ion source during long pulse operation

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

Hemsworth, R.S.; Boilson, D.; Crowley, B.

2006-03-15

It is foreseen to operate the neutral beam system of the International Thermonuclear Experimental Reactor (ITER) for pulse lengths extending up to 1 h. The performance of the KAMABOKO III negative ion source, which is a model of the source designed for ITER, is being studied on the MANTIS test bed at Cadarache. This article reports the latest results from the characterization of the ion source, in particular electron energy distribution measurements and the comparison between positive ion and negative ion extraction from the source.

Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

NASA Astrophysics Data System (ADS)

Borisov, A. A.; Deryabina, N. A.; Markovskij, D. V.

2017-12-01

Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing "expensive" experiments, is used for numerical integration of 235U and 238U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with 235U chambers and calibrate 238U chambers by responses of the 235U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor based on the average responses of the horizontal and vertical neutron chambers remains spatially stable as the source moves and can be used in addition to the staff monitor at neutron fluxes in the detectors four orders of magnitude lower than on the first wall, where staff detectors are located. Owing to low background, detectors of neutron chambers do not need calibration in the reactor because it is actually determination of the absolute detector efficiency for 14-MeV neutrons, which is a routine out-of-reactor procedure.

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. ...

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

WORKER STACKS GRAPHITE BLOCKS AGAINST INNER SOUTH WALL OF REACTOR. INL NEGATIVE NO. 3925. Unknown Photographer, 12/14/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

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

ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

ITER Status and Plans

NASA Astrophysics Data System (ADS)

Greenfield, Charles M.

2017-10-01

The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

Functional materials for breeding blankets—status and developments

NASA Astrophysics Data System (ADS)

Konishi, S.; Enoeda, M.; Nakamichi, M.; Hoshino, T.; Ying, A.; Sharafat, S.; Smolentsev, S.

2017-09-01

The development of tritium breeder, neutron multiplier and flow channel insert materials for the breeding blanket of the DEMO reactor is reviewed. Present emphasis is on the ITER test blanket module (TBM); lithium metatitanate (Li2TiO3) and lithium orthosilicate (Li4SiO4) pebbles have been developed by leading TBM parties. Beryllium pebbles have been selected as the neutron multiplier. Good progress has been made in their fabrication; however, verification of the design by experiments is in the planning stage. Irradiation data are also limited, but the decrease in thermal conductivity of beryllium due to irradiation followed by swelling is a concern. Tests at ITER are regarded as a major milestone. For the DEMO reactor, improvement of the breeder has been attempted to obtain a higher lithium content, and Be12Ti and other beryllide intermetallic compounds that have superior chemical stability have been studied. LiPb eutectic has been considered as a DEMO blanket in the liquid breeder option and is used as a coolant to achieve a higher outlet temperature; a SiC flow channel insert is used to prevent magnetohydrodynamic pressure drop and corrosion. A significant technical gap between ITER TBM and DEMO is recognized, and the world fusion community is working on ITER TBM and DEMO blanket development in parallel.

Progress on ion cyclotron range of frequencies heating physics and technology in support of the International Tokamak Experimental Reactor

NASA Astrophysics Data System (ADS)

Wilson, J. R.; Bonoli, P. T.

2015-02-01

Ion cyclotron range of frequency (ICRF) heating is foreseen as an integral component of the initial ITER operation. The status of ICRF preparations for ITER and supporting research were updated in the 2007 [Gormezano et al., Nucl. Fusion 47, S285 (2007)] report on the ITER physics basis. In this report, we summarize progress made toward the successful application of ICRF power on ITER since that time. Significant advances have been made in support of the technical design by development of new techniques for arc protection, new algorithms for tuning and matching, carrying out experimental tests of more ITER like antennas and demonstration on mockups that the design assumptions are correct. In addition, new applications of the ICRF system, beyond just bulk heating, have been proposed and explored.

GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

2015-09-01

An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

SNL-SAND-IV v. 0.9 (beta)

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

Griffin, Patrick J.

2016-10-05

The code is used to provide an unfolded/adjusted energy-dependent fission reactor neutron spectrum based upon an input trial spectrum and a set of measured activities. This is part of a neutron environment characterization that supports doing testing in a given reactor environment. An iterative perturbation method is used to obtain a "best fit" neutron flux spectrum for a given input set of infinitely dilute foil activities. The calculational procedure consists of the selection of a trial flux spectrum to serve as the initial approximation to the solution, and subsequent iteration to a form acceptable as an appropriate solution. The solutionmore » is specified either as time-integrated flux (fluence) for a pulsed environment or as a flux for a steady-state neutron environment.« less

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

ENGINEERING TEST REACTOR, TRA642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. ...

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

ENGINEERING TEST REACTOR, TRA-642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. CAMERA IS ON ROOF OF MTR BUILDING AND FACES DUE SOUTH. MTR SERVICE BUILDING, TRA-635, IN LOWER RIGHT CORNER. STEEL FRAMES SHOW BUILDINGS TO BE ATTACHED TO ETR BUILDING. HIGH-BAY SECTION IN CENTER IS REACTOR BUILDING. TWO-STORY CONTROL ROOM AND OFFICE BUILDING, TRA-647, IS BETWEEN IT AND MTR SERVICE BUILDING. STRUCTURE TO THE LEFT (WITH NO FRAMING YET) IS COMPRESSOR BUILDING, TRA-643, AND BEYOND IT WILL BE HEAT EXCHANGER BUILDING, TRA-644, GREAT SOUTHERN BUTTE ON HORIZON. INL NEGATIVE NO. 56-2382. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Astrophysics Data System (ADS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-09-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Technical Reports Server (NTRS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-01-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Space rocket engine on the base of the reactor-pumped laser for the interplanetary flights and earth orbital applications

NASA Astrophysics Data System (ADS)

Gulevich, Andrey V.; Dyachenko, Peter P.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.

2000-01-01

In this report the concept of vehicle-based reactor-laser engine for long time interplanetary and interorbital (LEO to GEO) flights is proposed. Reactor-pumped lasers offer the perspective way to create on the base of modern nuclear and lasers technologies the low mass and high energy density, repetitively pulsed vehicle-based laser of average power 100 kW. Nowadays the efficiency of nuclear-to-optical energy conversion reached the value of 2-3%. The demo model of reactor-pumped laser facility is under construction in Institute for Physics and Power Engineering (Obninsk, Russia). It enable us to hope that using high power laser on board of the vehicle could make the effective space laser engine possible. Such engine may provide the high specific impulse ~1000-2000 s with the thrust up to 10-100 n. Some calculation results of the characteristics of vehicle-based reactor-laser thermal engine concept are also presented. .

ETRCF, TRA654, INTERIOR. REACTOR OPERATED IN WATERFILLED TANK. CAMERA LOOKS ...

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

ETR-CF, TRA-654, INTERIOR. REACTOR OPERATED IN WATER-FILLED TANK. CAMERA LOOKS DOWN FROM ABOVE UPON LATER (NON-NUCLEAR) EXPERIMENTAL GEAR. INL NEGATIVE NO. HD24-1-1. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED ...

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

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED OUTSIDE OF MTR FOR EXPERIMENTS. THE AIRCRAFT NUCLEAR PROPULSION PROJECT DOMINATED THE USE OF THIS PART OF THE MTR. INL NEGATIVE NO. 7225. Unknown Photographer, 11/28/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Variable aperture-based ptychographical iterative engine method.

PubMed

Sun, Aihui; Kong, Yan; Meng, Xin; He, Xiaoliang; Du, Ruijun; Jiang, Zhilong; Liu, Fei; Xue, Liang; Wang, Shouyu; Liu, Cheng

2018-02-01

A variable aperture-based ptychographical iterative engine (vaPIE) is demonstrated both numerically and experimentally to reconstruct the sample phase and amplitude rapidly. By adjusting the size of a tiny aperture under the illumination of a parallel light beam to change the illumination on the sample step by step and recording the corresponding diffraction patterns sequentially, both the sample phase and amplitude can be faithfully reconstructed with a modified ptychographical iterative engine (PIE) algorithm. Since many fewer diffraction patterns are required than in common PIE and the shape, the size, and the position of the aperture need not to be known exactly, this proposed vaPIE method remarkably reduces the data acquisition time and makes PIE less dependent on the mechanical accuracy of the translation stage; therefore, the proposed technique can be potentially applied for various scientific researches. (2018) COPYRIGHT Society of Photo-Optical Instrumentation Engineers (SPIE).

Helping System Engineers Bridge the Peaks

NASA Technical Reports Server (NTRS)

Rungta, Neha; Tkachuk, Oksana; Person, Suzette; Biatek, Jason; Whalen, Michael W.; Castle, Joseph; Castle, JosephGundy-Burlet, Karen

2014-01-01

In our experience at NASA, system engineers generally follow the Twin Peaks approach when developing safety-critical systems. However, iterations between the peaks require considerable manual, and in some cases duplicate, effort. A significant part of the manual effort stems from the fact that requirements are written in English natural language rather than a formal notation. In this work, we propose an approach that enables system engineers to leverage formal requirements and automated test generation to streamline iterations, effectively "bridging the peaks". The key to the approach is a formal language notation that a) system engineers are comfortable with, b) is supported by a family of automated V&V tools, and c) is semantically rich enough to describe the requirements of interest. We believe the combination of formalizing requirements and providing tool support to automate the iterations will lead to a more efficient Twin Peaks implementation at NASA.

On the safety of ITER accelerators.

PubMed

Li, Ge

2013-01-01

Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

On the safety of ITER accelerators

PubMed Central

Li, Ge

2013-01-01

Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate −1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER. PMID:24008267

Iterative procedures for space shuttle main engine performance models

NASA Technical Reports Server (NTRS)

Santi, L. Michael

1989-01-01

Performance models of the Space Shuttle Main Engine (SSME) contain iterative strategies for determining approximate solutions to nonlinear equations reflecting fundamental mass, energy, and pressure balances within engine flow systems. Both univariate and multivariate Newton-Raphson algorithms are employed in the current version of the engine Test Information Program (TIP). Computational efficiency and reliability of these procedures is examined. A modified trust region form of the multivariate Newton-Raphson method is implemented and shown to be superior for off nominal engine performance predictions. A heuristic form of Broyden's Rank One method is also tested and favorable results based on this algorithm are presented.

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

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

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

2015-03-01

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

Burn Control in Fusion Reactors via Isotopic Fuel Tailoring

NASA Astrophysics Data System (ADS)

Boyer, Mark D.; Schuster, Eugenio

2011-10-01

The control of plasma density and temperature are among the most fundamental problems in fusion reactors and will be critical to the success of burning plasma experiments like ITER. Economic and technological constraints may require future commercial reactors to operate with low temperature, high-density plasma, for which the burn condition may be unstable. An active control system will be essential for stabilizing such operating points. In this work, a volume-averaged transport model for the energy and the densities of deuterium and tritium fuel ions, as well as the alpha particles, is used to synthesize a nonlinear feedback controller for stabilizing the burn condition. The controller makes use of ITER's planned isotopic fueling capability and controls the densities of these ions separately. The ability to modulate the DT fuel mix is exploited in order to reduce the fusion power during thermal excursions without the need for impurity injection. By moving the isotopic mix in the plasma away from the optimal 50:50 mix, the reaction rate is slowed and the alpha-particle heating is reduced to desired levels. Supported by the NSF CAREER award program (ECCS-0645086).

Polymerization Reactor Engineering.

ERIC Educational Resources Information Center

Skaates, J. Michael

1987-01-01

Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE ...

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

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE OF REACTOR. INL NEGATIVE NO. 4000. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Evaluation of inter-laminar shear strength of GFRP composed of bonded glass/polyimide tapes and cyanate-ester/epoxy blended resin for ITER TF coils

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

Hemmi, T.; Matsui, K.; Koizumi, N.

2014-01-27

The insulation system of the ITER TF coils consists of multi-layer glass/polyimide tapes impregnated a cyanate-ester/epoxy resin. The ITER TF coils are required to withstand an irradiation of 10 MGy from gamma-ray and neutrons since the ITER TF coils is exposed by fast neutron (>0.1 MeV) of 10{sup 22} n/m{sup 2} during the ITER operation. Cyanate-ester/epoxy blended resins and bonded glass/polyimide tapes are developed as insulation materials to realize the required radiation-hardness for the insulation of the ITER TF coils. To evaluate the radiation-hardness of the developed insulation materials, the inter-laminar shear strength (ILSS) of glass-fiber reinforced plastics (GFRP) fabricatedmore » using developed insulation materials is measured as one of most important mechanical properties before/after the irradiation in a fission reactor of JRR-3M. As a result, it is demonstrated that the GFRPs using the developed insulation materials have a sufficient performance to apply for the ITER TF coil insulation.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

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

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Engine System Model Development for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Nelson, Karl W.; Simpson, Steven P.

2006-01-01

In order to design, analyze, and evaluate conceptual Nuclear Thermal Propulsion (NTP) engine systems, an improved NTP design and analysis tool has been developed. The NTP tool utilizes the Rocket Engine Transient Simulation (ROCETS) system tool and many of the routines from the Enabler reactor model found in Nuclear Engine System Simulation (NESS). Improved non-nuclear component models and an external shield model were added to the tool. With the addition of a nearly complete system reliability model, the tool will provide performance, sizing, and reliability data for NERVA-Derived NTP engine systems. A new detailed reactor model is also being developed and will replace Enabler. The new model will allow more flexibility in reactor geometry and include detailed thermal hydraulics and neutronics models. A description of the reactor, component, and reliability models is provided. Another key feature of the modeling process is the use of comprehensive spreadsheets for each engine case. The spreadsheets include individual worksheets for each subsystem with data, plots, and scaled figures, making the output very useful to each engineering discipline. Sample performance and sizing results with the Enabler reactor model are provided including sensitivities. Before selecting an engine design, all figures of merit must be considered including the overall impacts on the vehicle and mission. Evaluations based on key figures of merit of these results and results with the new reactor model will be performed. The impacts of clustering and external shielding will also be addressed. Over time, the reactor model will be upgraded to design and analyze other NTP concepts with CERMET and carbide fuel cores.

Pollution Reduction Technology Program for Small Jet Aircraft Engines, Phase 2

NASA Technical Reports Server (NTRS)

Bruce, T. W.; Davis, F. G.; Kuhn, T. E.; Mongia, H. C.

1978-01-01

A series of iterative combustor pressure rig tests were conducted on two combustor concepts applied to the AiResearch TFE731-2 turbofan engine combustion system for the purpose of optimizing combustor performance and operating characteristics consistant with low emissions. The two concepts were an axial air-assisted airblast fuel injection configuration with variable-geometry air swirlers and a staged premix/prevaporization configuration. The iterative rig testing and modification sequence on both concepts was intended to provide operational compatibility with the engine and determine one concept for further evaluation in a TFE731-2 engine.

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

NASA Astrophysics Data System (ADS)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-12-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

REACTOR SERVICE BUILDING, TRA635. CROWDED MOCKUP AREA. CAMERA FACES EAST. ...

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

REACTOR SERVICE BUILDING, TRA-635. CROWDED MOCK-UP AREA. CAMERA FACES EAST. PHOTOGRAPHER'S NOTE SAYS "PICTURE REQUESTED BY IDO IN SUPPORT OF FY '58 BUILDING PROJECTS." INL NEGATIVE NO. 56-3025. R.G. Larsen, Photographer, 9/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Conceptual study of the cryocascade for pumping, separation and recycling of ITER torus exhaust

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

Mack, A.; Perinic, D.

1994-12-31

For pumping, separation and recycling of the ITER plasma exhaust, a pumping system working reliably under ambient and operating conditions of the ITER reactor is required. The pump is exposed to a magnetic field of about 3 T and has to be resistant to radioactive irradiation of 10{sup 9} rad. In the burn and dwell phase, a gas mixture consisting of hydrogen isotopes, helium ash and impurities has to be pumped at the pressure range of 10{sup {minus}2} - 10{sup {minus}1} mbar. Within the framework of the European Fusion Technology Programme, the concept of a primary cryopump for use inmore » ITER is being prepared at KfK. The cryocascade concept is planned to include three pump stages. These pump stages, which are connected in series, consist of individual chambers that may be separated from each other by means of cold valves. In the first stage, the impurities of the reactor exhaust gas are frozen out at 20-30 K. Settling of the hydrogen isotopes H/D/T on the 5 K cryosurfaces takes place in the second stage. This stage is made up of two parallel chambers, which can be switched from the pumping to the regeneration mode or vice versa. The helium fraction is bound in the downstream 5 K adsorption stage.« less

Nuclear Engine System Simulation (NESS) version 2.0

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

Optimizing Characterization of Site Hydrology in Support of New Reactor Licensing at the U.S. Nuclear Regulatory Commission (Invited)

NASA Astrophysics Data System (ADS)

Nicholson, T. J.; Raione, R.; Ahn, H.; Barnhurst, D.; Giacinto, J.; McBride, M.; Tiruneh, N. D.

2009-12-01

The NRC regulates the civilian use of radioactive materials and facilities in an open and transparent manner. The NRC regulatory criteria are designed to protect human health and safety, and the environment by regulating nuclear facilities. During review of new reactor licensing applications, NRC staff reviews and independently verifies hydrogeologic information submitted by the applicant in several topical areas such as development and testing of Conceptual Site Models (CSM) which may involve perched aquifers; engineered water level fluctuations of surface-water reservoirs; ground-water collector wells and local ground-water uses; design-basis ground-water levels for structural analysis; analysis of scenarios for potential release of radionuclides to the subsurface; deep well injection of effluents; and monitoring to detect radionuclide releases. This information is reviewed in a systematic manner in accordance with NRC requirements and guidance to evaluate safety and environmental impacts and reduce the uncertainties for these impacts. NRC licensing staff is reviewing 14 applications for siting new reactors. Experience gained through these licensing activities has shown the value of using site-specific data to evaluate the CSM and its use to assess design and operational issues. Optimizing the information flow process through a systemically and thorough review process creates efficiencies. Through an iterative process of evaluating various geographical settings and associated ground-water conditions, NRC staff has developed methods to minimize prediction uncertainty through the use of confirmatory analyses performed under conservative, hierarchal approaches.

Combustion engine system

NASA Technical Reports Server (NTRS)

Houseman, John (Inventor); Voecks, Gerald E. (Inventor)

1986-01-01

A flow through catalytic reactor which selectively catalytically decomposes methanol into a soot free hydrogen rich product gas utilizing engine exhaust at temperatures of 200 to 650 C to provide the heat for vaporizing and decomposing the methanol is described. The reactor is combined with either a spark ignited or compression ignited internal combustion engine or a gas turbine to provide a combustion engine system. The system may be fueled entirely by the hydrogen rich gas produced in the methanol decomposition reactor or the system may be operated on mixed fuels for transient power gain and for cold start of the engine system. The reactor includes a decomposition zone formed by a plurality of elongated cylinders which contain a body of vapor permeable, methanol decomposition catalyst preferably a shift catalyst such as copper-zinc.

Space Nuclear Reactor Engineering

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

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

NASA Astrophysics Data System (ADS)

Yu, Guanying; Liu, Xufeng; Liu, Songlin

2016-10-01

The Chinese Fusion Engineering Tokamak Reactor (CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder (WCCB) blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic (RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement. supported by National Natural Science Foundation of China (No. 11175207) and the National Magnetic Confinement Fusion Program of China (No. 2013GB108004)

ETR BUILDING, TRA642, INTERIOR. FIRST FLOOR. REACTOR IS IN CENTER ...

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

ETR BUILDING, TRA-642, INTERIOR. FIRST FLOOR. REACTOR IS IN CENTER OF VIEW. CAMERA FACES NORTHWEST. NOTE CRANE RAILS AND DANGLING ELECTRICAL CABLE AT UPPER PART OF VIEW FOR "MOFFETT 2 TON" CRANE. INL NEGATIVE NO. HD46-14-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED ...

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

ETR, TRA-642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED WITHIN THE INNER METAL FORM. WHEN CONCRETE IS POURED OUTSIDE THIS FORM, CONDUIT HOLES WILL BE PRESERVE SPACE THROUGH HOLES. INL NEGATIVE NO. 56-1507. Jack L. Anderson, Photographer, 5/8/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR ...

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

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR IS FUELED AS AN ETR MOCK-UP. LIGHTS DANGLE BELOW WATER LEVEL. CONTROL RODS AND OTHER APPARATUS DESCEND FROM ABOVE WATER LEVEL. INL NEGATIVE NO. 56-900. Jack L. Anderson, Photographer, 3/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR SERVICES BUILDING, TRA635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING ...

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

REACTOR SERVICES BUILDING, TRA-635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING AREA AND LABORATORY. CAMERA ON FIRST FLOOR FACING NORTH TOWARD MTR BUILDING. MOCK-UP AREA WAS TO THE RIGHT OF VIEW. INL NEGATIVE NO. HD46-10-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Nuclear engine flow reactivity shim control

DOEpatents

Walsh, J.M.

1973-12-11

A nuclear engine control system is provided which automatically compensates for reactor reactivity uncertainties at the start of life and reactivity losses due to core corrosion during the reactor life in gas-cooled reactors. The coolant gas flow is varied automatically by means of specially provided control apparatus so that the reactor control drums maintain a predetermined steady state position throughout the reactor life. This permits the reactor to be designed for a constant drum position and results in a desirable, relatively flat temperature profile across the core. (Official Gazette)

Mechanical Characterization of the Iter Mock-Up Insulation after Reactor Irradiation

NASA Astrophysics Data System (ADS)

Prokopec, R.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

2010-04-01

The ITER mock-up project was launched in order to demonstrate the feasibility of an industrial impregnation process using the new cyanate ester/epoxy blend. The mock-up simulates the TF winding pack cross section by a stainless steel structure with the same dimensions as the TF winding pack at a length of 1 m. It consists of 7 plates simulating the double pancakes, each of them is wrapped with glass fiber/Kapton sandwich tapes. After stacking the 7 plates, additional insulation layers are wrapped to simulate the ground insulation. This paper presents the results of the mechanical quality tests on the mock-up pancake insulation. Tensile and short beam shear specimens were cut from the plates extracted from the mock-up and tested at 77 K using a servo-hydraulic material testing device. All tests were repeated after reactor irradiation to a fast neutron fluence of 1×1022 m-2 (E>0.1 MeV). In order to simulate the pulsed operation of ITER, tension-tension fatigue measurements were performed in the load controlled mode. Initial results show a high mechanical strength as expected from the high number of thin glass fiber layers, and an excellent homogeneity of the material.

Designing magnetic systems for reliability

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

Heitzenroeder, P.J.

1991-01-01

Designing magnetic system is an iterative process in which the requirements are set, a design is developed, materials and manufacturing processes are defined, interrelationships with the various elements of the system are established, engineering analyses are performed, and fault modes and effects are studied. Reliability requires that all elements of the design process, from the seemingly most straightforward such as utilities connection design and implementation, to the most sophisticated such as advanced finite element analyses, receives a balanced and appropriate level of attention. D.B. Montgomery's study of magnet failures has shown that the predominance of magnet failures tend not tomore » be in the most intensively engineered areas, but are associated with insulation, leads, ad unanticipated conditions. TFTR, JET, JT-60, and PBX are all major tokamaks which have suffered loss of reliability due to water leaks. Similarly the majority of causes of loss of magnet reliability at PPPL has not been in the sophisticated areas of the design but are due to difficulties associated with coolant connections, bus connections, and external structural connections. Looking towards the future, the major next-devices such as BPX and ITER are most costly and complex than any of their predecessors and are pressing the bounds of operating levels, materials, and fabrication. Emphasis on reliability is a must as the fusion program enters a phase where there are fewer, but very costly devices with the goal of reaching a reactor prototype stage in the next two or three decades. This paper reviews some of the magnet reliability issues which PPPL has faced over the years the lessons learned from them, and magnet design and fabrication practices which have been found to contribute to magnet reliability.« less

Reactor Operations Monitoring System

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

Hart, M.M.

1989-01-01

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

Formation and sustainment of internal transport barriers in the International Thermonuclear Experimental Reactor with the baseline heating mixa)

NASA Astrophysics Data System (ADS)

Poli, Francesca M.; Kessel, Charles E.

2013-05-01

Plasmas with internal transport barriers (ITBs) are a potential and attractive route to steady-state operation in ITER. These plasmas exhibit radially localized regions of improved confinement with steep pressure gradients in the plasma core, which drive large bootstrap current and generate hollow current profiles and negative magnetic shear. This work examines the formation and sustainment of ITBs in ITER with electron cyclotron heating and current drive. The time-dependent transport simulations indicate that, with a trade-off of the power delivered to the equatorial and to the upper launcher, the sustainment of steady-state ITBs can be demonstrated in ITER with the baseline heating configuration.

Variable aperture-based ptychographical iterative engine method

NASA Astrophysics Data System (ADS)

Sun, Aihui; Kong, Yan; Meng, Xin; He, Xiaoliang; Du, Ruijun; Jiang, Zhilong; Liu, Fei; Xue, Liang; Wang, Shouyu; Liu, Cheng

2018-02-01

A variable aperture-based ptychographical iterative engine (vaPIE) is demonstrated both numerically and experimentally to reconstruct the sample phase and amplitude rapidly. By adjusting the size of a tiny aperture under the illumination of a parallel light beam to change the illumination on the sample step by step and recording the corresponding diffraction patterns sequentially, both the sample phase and amplitude can be faithfully reconstructed with a modified ptychographical iterative engine (PIE) algorithm. Since many fewer diffraction patterns are required than in common PIE and the shape, the size, and the position of the aperture need not to be known exactly, this proposed vaPIE method remarkably reduces the data acquisition time and makes PIE less dependent on the mechanical accuracy of the translation stage; therefore, the proposed technique can be potentially applied for various scientific researches.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

NASA Astrophysics Data System (ADS)

Berg, Thomas A.; Disney, Richard K.

2004-02-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

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

Berg, Thomas A.; Disney, Richard K.

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

ENGINEERING APPLICATIONS OF ANALOG COMPUTERS

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

Bryant, L.T.; Janicke, M.J.; Just, L.C.

1963-10-31

Six experiments from the fields of reactor engineering, heat transfer, and dynamics are presented to illustrate the engineering applications of analog computers. The steps required for producing the analog solution are shown, as well as complete information for duplicating the solution. Graphical results are provided. The experiments include: deceleration of a reactor control rod, pressure variations through a packed bed, reactor kinetics over many decades with thermal feedback, a vibrating system with two degrees of freedom, temperature distribution in a radiating fin, temperature distribution in an infinite slab considering variable thermal properties, and iodine -xenon buildup in a reactor. (M.C.G.)

Comparative studies for two different orientations of pebble bed in an HCCB blanket

NASA Astrophysics Data System (ADS)

Paritosh, CHAUDHURI; Chandan, DANANI; E, RAJENDRAKUMAR

2017-12-01

The Indian Test Blanket Module (TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development (R&D) is focused on two types of breeding blanket concepts: lead-lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic R&D program for DEMO relevant technology development. In the HCCB concept Li2TiO3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept (case-1), the ceramic breeder beds are kept horizontal in the toroidal-radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept (case-2), the pebble bed is vertically (poloidal-radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations. Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.

Present limits and improvements of structural materials for fusion reactors - a review

NASA Astrophysics Data System (ADS)

Tavassoli, A.-A. F.

2002-04-01

Since the transition from ITER or DEMO to a commercial power reactor would involve a significant change in system and materials options, a parallel R&D path has been put in place in Europe to address these issues. This paper assesses the structural materials part of this program along with the latest R&D results from the main programs. It is shown that stainless steels and ferritic/martensitic steels, retained for ITER and DEMO, will also remain the principal contenders for the future FPR, despite uncertainties over irradiation induced embrittlement at low temperatures and consequences of high He/dpa ratio. Neither one of the present advanced high temperature materials has to this date the structural integrity reliability needed for application in critical components. This situation is unlikely to change with the materials R&D alone and has to be mitigated in close collaboration with blanket system design.

An assessment of coupling algorithms for nuclear reactor core physics simulations

DOE PAGES

Hamilton, Steven; Berrill, Mark; Clarno, Kevin; ...

2016-04-01

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

An assessment of coupling algorithms for nuclear reactor core physics simulations

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

Hamilton, Steven; Berrill, Mark; Clarno, Kevin

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

An assessment of coupling algorithms for nuclear reactor core physics simulations

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

Hamilton, Steven, E-mail: hamiltonsp@ornl.gov; Berrill, Mark, E-mail: berrillma@ornl.gov; Clarno, Kevin, E-mail: clarnokt@ornl.gov

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNKmore » and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

NASA Astrophysics Data System (ADS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Hamada, K.; Sugimoto, M.; Okuno, K.

2002-12-01

The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2×10 22 m -2 ( E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite.

Extrapolation techniques applied to matrix methods in neutron diffusion problems

NASA Technical Reports Server (NTRS)

Mccready, Robert R

1956-01-01

A general matrix method is developed for the solution of characteristic-value problems of the type arising in many physical applications. The scheme employed is essentially that of Gauss and Seidel with appropriate modifications needed to make it applicable to characteristic-value problems. An iterative procedure produces a sequence of estimates to the answer; and extrapolation techniques, based upon previous behavior of iterants, are utilized in speeding convergence. Theoretically sound limits are placed on the magnitude of the extrapolation that may be tolerated. This matrix method is applied to the problem of finding criticality and neutron fluxes in a nuclear reactor with control rods. The two-dimensional finite-difference approximation to the two-group neutron fluxes in a nuclear reactor with control rods. The two-dimensional finite-difference approximation to the two-group neutron-diffusion equations is treated. Results for this example are indicated.

Installation and Testing of ITER Integrated Modeling and Analysis Suite (IMAS) on DIII-D

NASA Astrophysics Data System (ADS)

Lao, L.; Kostuk, M.; Meneghini, O.; Smith, S.; Staebler, G.; Kalling, R.; Pinches, S.

2017-10-01

A critical objective of the ITER Integrated Modeling Program is the development of IMAS to support ITER plasma operation and research activities. An IMAS framework has been established based on the earlier work carried out within the EU. It consists of a physics data model and a workflow engine. The data model is capable of representing both simulation and experimental data and is applicable to ITER and other devices. IMAS has been successfully installed on a local DIII-D server using a flexible installer capable of managing the core data access tools (Access Layer and Data Dictionary) and optionally the Kepler workflow engine and coupling tools. A general adaptor for OMFIT (a workflow engine) is being built for adaptation of any analysis code to IMAS using a new IMAS universal access layer (UAL) interface developed from an existing OMFIT EU Integrated Tokamak Modeling UAL. Ongoing work includes development of a general adaptor for EFIT and TGLF based on this new UAL that can be readily extended for other physics codes within OMFIT. Work supported by US DOE under DE-FC02-04ER54698.

Nuclear Engineering Technologists in the Nuclear Power Era

ERIC Educational Resources Information Center

Wang, C. H.; And Others

1974-01-01

Describes manpower needs in nuclear engineering in the areas of research and development, architectural engineering and construction supervision, power reactor operations, and regulatory tasks. Outlines a suitable curriculum to prepare students for the tasks related to construction and operation of power reactors. (GS)

Semiconductor Chemical Reactor Engineering and Photovoltaic Unit Operations.

ERIC Educational Resources Information Center

Russell, T. W. F.

1985-01-01

Discusses the nature of semiconductor chemical reactor engineering, illustrating the application of this engineering with research in physical vapor deposition of cadmium sulfide at both the laboratory and unit operations scale and chemical vapor deposition of amorphous silicon at the laboratory scale. (JN)

Evolutionary engineering for industrial microbiology.

PubMed

Vanee, Niti; Fisher, Adam B; Fong, Stephen S

2012-01-01

Superficially, evolutionary engineering is a paradoxical field that balances competing interests. In natural settings, evolution iteratively selects and enriches subpopulations that are best adapted to a particular ecological niche using random processes such as genetic mutation. In engineering desired approaches utilize rational prospective design to address targeted problems. When considering details of evolutionary and engineering processes, more commonality can be found. Engineering relies on detailed knowledge of the problem parameters and design properties in order to predict design outcomes that would be an optimized solution. When detailed knowledge of a system is lacking, engineers often employ algorithmic search strategies to identify empirical solutions. Evolution epitomizes this iterative optimization by continuously diversifying design options from a parental design, and then selecting the progeny designs that represent satisfactory solutions. In this chapter, the technique of applying the natural principles of evolution to engineer microbes for industrial applications is discussed to highlight the challenges and principles of evolutionary engineering.

Optimal startup control of a jacketed tubular reactor.

NASA Technical Reports Server (NTRS)

Hahn, D. R.; Fan, L. T.; Hwang, C. L.

1971-01-01

The optimal startup policy of a jacketed tubular reactor, in which a first-order, reversible, exothermic reaction takes place, is presented. A distributed maximum principle is presented for determining weak necessary conditions for optimality of a diffusional distributed parameter system. A numerical technique is developed for practical implementation of the distributed maximum principle. This involves the sequential solution of the state and adjoint equations, in conjunction with a functional gradient technique for iteratively improving the control function.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Human factors engineering verification and validation for APR1400 computerized control room

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

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

2006-07-01

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

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

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

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

ECOMAT INC. BIOLOGICAL DENITRIFICATION PROCESS, ITER

EPA Science Inventory

EcoMat, Inc. of Hayward, California (EcoMat) has developed an ex situ anoxic biofilter biodenitrification (BDN) process. The process uses specific biocarriers and bacteria to treat nitrate-contaminated water and employs a patented reactor that retains biocarrier within the syste...

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE ...

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

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE OF ETR REACTOR, CAMERA FACING NORTH. CABINET CONTAINING "NUCLEAR INSTRUMENT SYSTEMS" IS RESTRICTED. INL NEGATIVE NO. HD46-18-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

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

ITER EDA Newsletter. Volume 3, no. 2

NASA Astrophysics Data System (ADS)

1994-02-01

This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the Fifth ITER Council Meeting held in Garching, Germany, January 27-28, 1994, a visit (January 28, 1994) of an international group of Harvard Fellows to the San Diego Joint Work Site, the Inauguration Ceremony of the EC-hosted ITER joint work site in Garching (January 28, 1994), on an ITER Technical Meeting on Assembly and Maintenance held in Garching, Germany, January 19-26, 1994, and a report on a Technical Committee Meeting on radiation effects on in-vessel components held in Garching, Germany, November 15-19, 1993, as well as an ITER Status Report.

REACTOR SERVICE BUILDING, TRA635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ...

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

REACTOR SERVICE BUILDING, TRA-635, CONTEXTUAL VIEW DURING CONSTRUCTION. CAMERA IS ATOP MTR BUILDING AND LOOKING SOUTHERLY. FOUNDATION AND DRAINS ARE UNDER CONSTRUCTION. THE BUILDING WILL BUTT AGAINST CHARGING FACE OF PLUG STORAGE BUILDING. HOT CELL BUILDING, TRA-632, IS UNDER CONSTRUCTION AT TOP CENTER OF VIEW. INL NEGATIVE NO. 8518. Unknown Photographer, 8/25/1953 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Applying chemical engineering concepts to non-thermal plasma reactors

NASA Astrophysics Data System (ADS)

Pedro AFFONSO, NOBREGA; Alain, GAUNAND; Vandad, ROHANI; François, CAUNEAU; Laurent, FULCHERI

2018-06-01

Process scale-up remains a considerable challenge for environmental applications of non-thermal plasmas. Undersanding the impact of reactor hydrodynamics in the performance of the process is a key step to overcome this challenge. In this work, we apply chemical engineering concepts to analyse the impact that different non-thermal plasma reactor configurations and regimes, such as laminar or plug flow, may have on the reactor performance. We do this in the particular context of the removal of pollutants by non-thermal plasmas, for which a simplified model is available. We generalise this model to different reactor configurations and, under certain hypotheses, we show that a reactor in the laminar regime may have a behaviour significantly different from one in the plug flow regime, often assumed in the non-thermal plasma literature. On the other hand, we show that a packed-bed reactor behaves very similarly to one in the plug flow regime. Beyond those results, the reader will find in this work a quick introduction to chemical reaction engineering concepts.

Evaluation of some candidate materials for automobile thermal reactors in engine-dynamometer screening tests

NASA Technical Reports Server (NTRS)

Oldrieve, R. E.

1971-01-01

Fourteen materials were evaluated in engine screening tests on full-size thermal reactors for automobile engine pollution control systems. Cyclic test-stand engine operation provided 2 hours at 1040 C and a 20-minute air-cool to 70 C each test cycle. Each reactor material was exposed to 83 cycles in 200 hours of engine testing. On the basis of resistance to oxidation and distortion, the best materials included two ferritic iron alloys (Ge 1541 and Armco 18S/R), several commercial oxidation-resistant coatings on AlSl 651 (19-9 DL), and possibly uncoated AISI 310. The best commercial coatings were Cr-Al, Ni-Cr, and a glass ceramic.

Carbon fiber composites application in ITER plasma facing components

NASA Astrophysics Data System (ADS)

Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

1998-10-01

Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

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

Merrill, Brad J

This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allowsmore » the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.« less

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

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

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR ...

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

TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR AREAS SOUTH OF PERCH AVENUE. "COLD" SERVICES NORTH OF PERCH. ADVANCED TEST REACTOR IN NEW SECTION WEST OF COLD SERVICES SECTION. NEW PERIMETER FENCE ENCLOSES BETA RAY SPECTROMETER, TRA-669, AN ATR SUPPORT FACILITY, AND ATR STACK. UTM LOCATORS HAVE BEEN DELETED. IDAHO NUCLEAR CORPORATION, FROM A BLAW-KNOX DRAWING, 3/1968. INL INDEX NO. 530-0100-00-400-011646, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Engine management during NTRE start up

NASA Technical Reports Server (NTRS)

Bulman, Mel; Saltzman, Dave

1993-01-01

The topics are presented in viewgraph form and include the following: total engine system management critical to successful nuclear thermal rocket engine (NTRE) start up; NERVA type engine start windows; reactor power control; heterogeneous reactor cooling; propellant feed system dynamics; integrated NTRE start sequence; moderator cooling loop and efficient NTRE starting; analytical simulation and low risk engine development; accurate simulation through dynamic coupling of physical processes; and integrated NTRE and mission performance.

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

Stimpson, Shane; Collins, Benjamin; Kochunas, Brendan

The MPACT code, being developed collaboratively by the University of Michigan and Oak Ridge National Laboratory, is the primary deterministic neutron transport solver being deployed within the Virtual Environment for Reactor Applications (VERA) as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL). In many applications of the MPACT code, transport-corrected scattering has proven to be an obstacle in terms of stability, and considerable effort has been made to try to resolve the convergence issues that arise from it. Most of the convergence problems seem related to the transport-corrected cross sections, particularly when used in the 2Dmore » method of characteristics (MOC) solver, which is the focus of this work. Here in this paper, the stability and performance of the 2-D MOC solver in MPACT is evaluated for two iteration schemes: Gauss-Seidel and Jacobi. With the Gauss-Seidel approach, as the MOC solver loops over groups, it uses the flux solution from the previous group to construct the inscatter source for the next group. Alternatively, the Jacobi approach uses only the fluxes from the previous outer iteration to determine the inscatter source for each group. Consequently for the Jacobi iteration, the loop over groups can be moved from the outermost loop$-$as is the case with the Gauss-Seidel sweeper$-$to the innermost loop, allowing for a substantial increase in efficiency by minimizing the overhead of retrieving segment, region, and surface index information from the ray tracing data. Several test problems are assessed: (1) Babcock & Wilcox 1810 Core I, (2) Dimple S01A-Sq, (3) VERA Progression Problem 5a, and (4) VERA Problem 2a. The Jacobi iteration exhibits better stability than Gauss-Seidel, allowing for converged solutions to be obtained over a much wider range of iteration control parameters. Additionally, the MOC solve time with the Jacobi approach is roughly 2.0-2.5× faster per sweep. While the performance and stability of the Jacobi iteration are substantially improved compared to the Gauss-Seidel iteration, it does yield a roughly 8$-$10% increase in the overall memory requirement.« less

Improvement of transport-corrected scattering stability and performance using a Jacobi inscatter algorithm for 2D-MOC

DOE PAGES

Stimpson, Shane; Collins, Benjamin; Kochunas, Brendan

2017-03-10

The MPACT code, being developed collaboratively by the University of Michigan and Oak Ridge National Laboratory, is the primary deterministic neutron transport solver being deployed within the Virtual Environment for Reactor Applications (VERA) as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL). In many applications of the MPACT code, transport-corrected scattering has proven to be an obstacle in terms of stability, and considerable effort has been made to try to resolve the convergence issues that arise from it. Most of the convergence problems seem related to the transport-corrected cross sections, particularly when used in the 2Dmore » method of characteristics (MOC) solver, which is the focus of this work. Here in this paper, the stability and performance of the 2-D MOC solver in MPACT is evaluated for two iteration schemes: Gauss-Seidel and Jacobi. With the Gauss-Seidel approach, as the MOC solver loops over groups, it uses the flux solution from the previous group to construct the inscatter source for the next group. Alternatively, the Jacobi approach uses only the fluxes from the previous outer iteration to determine the inscatter source for each group. Consequently for the Jacobi iteration, the loop over groups can be moved from the outermost loop$-$as is the case with the Gauss-Seidel sweeper$-$to the innermost loop, allowing for a substantial increase in efficiency by minimizing the overhead of retrieving segment, region, and surface index information from the ray tracing data. Several test problems are assessed: (1) Babcock & Wilcox 1810 Core I, (2) Dimple S01A-Sq, (3) VERA Progression Problem 5a, and (4) VERA Problem 2a. The Jacobi iteration exhibits better stability than Gauss-Seidel, allowing for converged solutions to be obtained over a much wider range of iteration control parameters. Additionally, the MOC solve time with the Jacobi approach is roughly 2.0-2.5× faster per sweep. While the performance and stability of the Jacobi iteration are substantially improved compared to the Gauss-Seidel iteration, it does yield a roughly 8$-$10% increase in the overall memory requirement.« less

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

HOT CELL BUILDING, TRA632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA ...

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

HOT CELL BUILDING, TRA-632. CONTEXTUAL VIEW ALONG WALLEYE AVENUE, CAMERA FACING EASTERLY. HOT CELL BUILDING IS AT CENTER LEFT OF VIEW; THE LOW-BAY PROJECTION WITH LADDER IS THE TEST TRAIN ASSEMBLY FACILITY, ADDED IN 1968. MTR BUILDING IS IN LEFT OF VIEW. HIGH-BAY BUILDING AT RIGHT IS THE ENGINEERING TEST REACTOR BUILDING, TRA-642. INL NEGATIVE NO. HD46-32-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

NASA Technical Reports Server (NTRS)

Sorensen, Kirk; Juhasz, Albert

2007-01-01

Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

A superlinear interior points algorithm for engineering design optimization

NASA Technical Reports Server (NTRS)

Herskovits, J.; Asquier, J.

1990-01-01

We present a quasi-Newton interior points algorithm for nonlinear constrained optimization. It is based on a general approach consisting of the iterative solution in the primal and dual spaces of the equalities in Karush-Kuhn-Tucker optimality conditions. This is done in such a way to have primal and dual feasibility at each iteration, which ensures satisfaction of those optimality conditions at the limit points. This approach is very strong and efficient, since at each iteration it only requires the solution of two linear systems with the same matrix, instead of quadratic programming subproblems. It is also particularly appropriate for engineering design optimization inasmuch at each iteration a feasible design is obtained. The present algorithm uses a quasi-Newton approximation of the second derivative of the Lagrangian function in order to have superlinear asymptotic convergence. We discuss theoretical aspects of the algorithm and its computer implementation.

Thermal modeling of W rod armor.

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

Nygren, Richard Einar

2004-09-01

Sandia has developed and tested mockups armored with W rods over the last decade and pioneered the initial development of W rod armor for International Thermonuclear Experimental Reactor (ITER) in the 1990's. We have also developed 2D and 3D thermal and stress models of W rod-armored plasma facing components (PFCs) and test mockups and are applying the models to both short pulses, i.e. edge localized modes (ELMs), and thermal performance in steady state for applications in C-MOD, DiMES testing and ITER. This paper briefly describes the 2D and 3D models and their applications with emphasis on modeling for an ongoingmore » test program that simulates repeated heat loads from ITER ELMs.« less

Researcher Poses with a Nuclear Rocket Model

NASA Image and Video Library

1961-11-21

A researcher at the NASA Lewis Research Center with slide ruler poses with models of the earth and a nuclear-propelled rocket. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The nuclear rocket model in this photograph includes a reactor at the far right with a hydrogen propellant tank and large radiator below. The payload or crew would be at the far left, distanced from the reactor.

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

Exploratory evaluation of ceramics for automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1972-01-01

An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

Conceptual Design and Analysis of Cold Mass Support of the CS3U Feeder for the ITER

NASA Astrophysics Data System (ADS)

Zhu, Yinfeng; Song, Yuntao; Zhang, Yuanbin; Wang, Zhongwei

2013-06-01

In the International Thermonuclear Experimental Reactor (ITER) project, the feeders are one of the most important and critical systems. To convey the power supply and the coolant for the central solenoid (CS) magnet, 6 sets of CS feeders are employed, which consist mainly of an in-cryostat feeder (ICF), a cryostat feed-through (CFT), an S-bend box (SBB), and a coil terminal box (CTB). To compensate the displacements of the internal components of the CS feeders during operation, sliding cold mass supports consisting of a sled plate, a cylindrical support, a thermal shield, and an external ring are developed. To check the strength of the developed cold mass supports of the CS3U feeder, electromagnetic analysis of the two superconducting busbars is performed by using the CATIA V5 and ANSYS codes based on parametric technology. Furthermore, the thermal-structural coupling analysis is performed based on the obtained results, except for the stress concentration, and the max. stress intensity is lower than the allowable stress of the selected material. It is found that the conceptual design of the cold mass support can satisfy the required functions under the worst case of normal working conditions. All these performed activities will provide a firm technical basis for the engineering design and development of cold mass supports.

A Simple Experiment for Teaching Process Intensification by Static Mixing in Chemical Reaction Engineering

ERIC Educational Resources Information Center

Baz-Rodríguez, Sergio; Herrera-Soberanis, Natali; Rodríguez-Novelo, Miguel; Guillén-Francisc, Juana; Rocha-Uribe, José

2016-01-01

An experiment for teaching mixing intensification in reaction engineering is described. For this, a simple tubular reactor was constructed; helical static mixer elements were fabricated from stainless steel strips and inserted into the reactor. With and without the internals, the equipment operates as a static mixer reactor or a laminar flow…

ETR, TRA642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN ...

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

ETR, TRA-642. ON GROUND FLOOR. WITH OUTER THERMAL RING IN PLACE AND CONDUIT PRESERVED, HIGH-DENSITY CONCRETE IS PLACED BETWEEN THE THERMAL RING AND THE OUTER REACTOR FORM. INL NEGATIVE NO. 56-2400. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION ...

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

EAST FACE OF REACTOR BASE. COMING TOWARD CAMERA IS EXCAVATION FOR MTR CANAL. CAISSONS FLANK EACH SIDE. COUNTERFORT (SUPPORT PERPENDICULAR TO WHAT WILL BE THE LONG WALL OF THE CANAL) RESTS ATOP LEFT CAISSON. IN LOWER PART OF VIEW, DRILLERS PREPARE TRENCHES FOR SUPPORT BEAMS THAT WILL LIE BENEATH CANAL FLOOR. INL NEGATIVE NO. 739. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ON GROUND FLOOR. THE 60TON ETR REACTOR VESSEL ...

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

ETR, TRA-642. ON GROUND FLOOR. THE 60-TON ETR REACTOR VESSEL IS DROPPED INTO PLACE OVER PIT. KAISER USED TWO MULTI-BLOCK DRUM PULLEYS WITH A COMBINED CAPACITY OF 100 TONS AND A 100-TON DRUM HOIST. THE VESSEL WAS 35 1/2 FEET LONG. INL NEGATIVE NO. 56-4149. R.G. Larsen, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

PubMed

Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

2018-03-01

In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

IDC Re-Engineering Phase 2 Iteration E2 Use Case Realizations Version 1.2.

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

Hamlet, Benjamin R.; Harris, James M.; Burns, John F.

2016-12-01

This document contains 4 use case realizations generated from the model contained in Rational Software Architect. These use case realizations are the current versions of the realizations originally delivered in Elaboration Iteration 2.

Generic Stellarator-like Magnetic Fusion Reactor

NASA Astrophysics Data System (ADS)

Sheffield, John; Spong, Donald

2015-11-01

The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

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

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

Ambrose, T.W.

1965-06-04

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

Beryllium fabrication/cost assessment for ITER (International Thermonuclear Experimental Reactor)

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

Beeston, J.M.; Longhurst, G.R.; Parsonage, T.

1990-06-01

A fabrication and cost estimate of three possible beryllium shapes for the International Thermonuclear Experimental Reactor (ITER) blanket is presented. The fabrication method by hot pressing (HP), cold isostatic pressing plus sintering (CIP+S), cold isostatic pressing plus sintering plus hot isostatic pressing (CIP+S+HIP), and sphere production by atomization or rotary electrode will be discussed. Conventional hot pressing blocks of beryllium with subsequent machining to finished shapes can be more expensive than production of a net shape by cold isostatic pressing and sintering. The three beryllium shapes to be considered here and proposed for ITER are: (1) cubic blocks (3 tomore » 17 cm on an edge), (2) tubular cylinders (33 to 50 mm i.d. by 62 mm o.d. by 8 m long), and (3) spheres (1--5 mm dia.). A rough cost estimate of the basic shape is presented which would need to be refined if the surface finish and tolerances required are better than the sintering process produces. The final cost of the beryllium in the blanket will depend largely on the machining and recycling of beryllium required to produce the finished product. The powder preparation will be discussed before shape fabrication. 10 refs., 6 figs.« less

Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

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

Zhai, Y.; Loesser, G.; Smith, M.

ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses weremore » performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.« less

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

An evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were analyzed in cyclic engine dynamometer tests with peak temperature of 1900 F (1040 C). Two developmental ferritic iron alloys GE1541 and NASA-18T - exhibited the best overall performance lasting at least 60% of the life of the test engine. Four of the alloys evaluated warrant consideration for reactor use. They include GE1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.-

Evaluation of alloys and coatings for use in automobile thermal reactors

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Oldrieve, R. E.

1974-01-01

Several candidate alloys and coatings were evaluated for use in automobile thermal reactors. Full-size reactors of the candidate materials were evaluated in cyclic engine dynamometer tests with a peak temperature of 1040 C (1900 F). Two developmental ferritic-iron alloys, GE-1541 and NASA-18T, exhibited the best overall performance by lasting at least 60 percent of the life of test engine. Four of the alloys evaluated warrant consideration for reactor use. They are GE-1541, Armco 18 SR, NASA-18T, and Inconel 601. None of the commercial coating substrate combinations evaluated warrant consideration for reactor use.

MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

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

MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR SERVICE BUILDING, TRA635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR ...

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

REACTOR SERVICE BUILDING, TRA-635, INTERIOR. CAMERA FACES NORTHWEST TOWARDS INTERIOR WALL ENCLOSING STORAGE AND OFFICE SPACE ALONG THE WEST SIDE. AT RIGHT EDGE IS DOOR TO MTR BUILDING. FROM RIGHT TO LEFT, SPACE WAS PLANNED FOR A LOCKER ROOM, MTR ISSUE ROOM, AND STORAGE AREAS AND RELATED OFFICES. NOTE SECOND "MEZZANINE" FLOOR ABOVE. INL NEGATIVE NO. 10227. Unknown Photographer, 3/23/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETRCF, TRA654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE ...

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

ETR-CF, TRA-654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE HOOKS. ELECTRICAL EQUIPMENT IS PART OF PAST EXPERIMENT. DOOR AT LEFT EDGE OF VIEW LEADS TO REACTOR SERVICE BUILDING, TRA-635. INL NEGATIVE NO. HD24-1-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

ERIC Educational Resources Information Center

Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

2014-01-01

An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

Physics and nuclear power

NASA Astrophysics Data System (ADS)

Buttery, N. E.

2008-03-01

Nuclear power owes its origin to physicists. Fission was demonstrated by physicists and chemists and the first nuclear reactor project was led by physicists. However as nuclear power was harnessed to produce electricity the role of the engineer became stronger. Modern nuclear power reactors bring together the skills of physicists, chemists, chemical engineers, electrical engineers, mechanical engineers and civil engineers. The paper illustrates this by considering the Sizewell B project and the role played by physicists in this. This covers not only the roles in design and analysis but in problem solving during the commissioning of first of a kind plant. Looking forward to the challenges to provide sustainable and environmentally acceptable energy sources for the future illustrates the need for a continuing synergy between physics and engineering. This will be discussed in the context of the challenges posed by Generation IV reactors.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

ETR ELECTRICAL BUILDING, TRA648, INTERIOR. SWITCHGEAR. INL NEGATIVE NO. 563794. ...

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

ETR ELECTRICAL BUILDING, TRA-648, INTERIOR. SWITCHGEAR. INL NEGATIVE NO. 56-3794. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. BATTERY ROOM. INL NEGATIVE NO. 563785. ...

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

ETR ELECTRICAL BUILDING, TRA-648. BATTERY ROOM. INL NEGATIVE NO. 56-3785. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. INSIDE A FLASH EVAPORATOR. INL NEGATIVE ...

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

PROCESS WATER BUILDING, TRA-605. INSIDE A FLASH EVAPORATOR. INL NEGATIVE NO. 3323. Unknown Photographer, 9/12/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Low-temperature tensile strength of the ITER-TF model coil insulation system after reactor irradiation

NASA Astrophysics Data System (ADS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.

The windings of the superconducting magnet coils for the ITER-FEAT fusion device are affected by high mechanical stresses at cryogenic temperatures and by a radiation environment, which impose certain constraints especially on the insulating materials. A glass fiber reinforced plastic (GFRP) laminate, which consists of Kapton/R-glass-fiber reinforcement tapes, vacuum-impregnated in a DGEBA epoxy system, was used for the European toroidal field model coil turn insulation of ITER. In order to assess its mechanical properties under the actual operating conditions of ITER-FEAT, cryogenic (77 K) static tensile tests and tension-tension fatigue measurements were done before and after irradiation to a fast neutron fluence of 1×10 22 m -2 ( E>0.1 MeV), i.e. the ITER-FEAT design fluence level. We find that the mechanical strength and the fracture behavior of this GFRP are strongly influenced by the winding direction of the tape and by the radiation induced delamination process. In addition, the composite swells by 3%, forming bubbles inside the laminate, and loses weight (1.4%) at the design fluence.

A stopping criterion to halt iterations at the Richardson-Lucy deconvolution of radiographic images

NASA Astrophysics Data System (ADS)

Almeida, G. L.; Silvani, M. I.; Souza, E. S.; Lopes, R. T.

2015-07-01

Radiographic images, as any experimentally acquired ones, are affected by spoiling agents which degrade their final quality. The degradation caused by agents of systematic character, can be reduced by some kind of treatment such as an iterative deconvolution. This approach requires two parameters, namely the system resolution and the best number of iterations in order to achieve the best final image. This work proposes a novel procedure to estimate the best number of iterations, which replaces the cumbersome visual inspection by a comparison of numbers. These numbers are deduced from the image histograms, taking into account the global difference G between them for two subsequent iterations. The developed algorithm, including a Richardson-Lucy deconvolution procedure has been embodied into a Fortran program capable to plot the 1st derivative of G as the processing progresses and to stop it automatically when this derivative - within the data dispersion - reaches zero. The radiograph of a specially chosen object acquired with thermal neutrons from the Argonauta research reactor at Institutode Engenharia Nuclear - CNEN, Rio de Janeiro, Brazil, have undergone this treatment with fair results.

Incorporating Multi-criteria Optimization and Uncertainty Analysis in the Model-Based Systems Engineering of an Autonomous Surface Craft

DTIC Science & Technology

2009-09-01

SAS Statistical Analysis Software SE Systems Engineering SEP Systems Engineering Process SHP Shaft Horsepower SIGINT Signals Intelligence......management occurs (OSD 2002). The Systems Engineering Process (SEP), displayed in Figure 2, is a comprehensive , iterative and recursive problem

Dielectric strength, swelling and weight loss of the ITER Toroidal Field Model Coil insulation after low temperature reactor irradiation

NASA Astrophysics Data System (ADS)

Humer, K.; Weber, H. W.; Hastik, R.; Hauser, H.; Gerstenberg, H.

2000-04-01

The insulation system for the Toroidal Field Model Coil of ITER is a fiber reinforced plastic (FRP) laminate, which consists of a combined Kapton/R-glass-fiber reinforcement tape, vacuum-impregnated with an epoxy DGEBA system. Pure disk shaped laminates, FRP/stainless-steel sandwiches, and conductor insulation prototypes were irradiated at 5 K in a fission reactor up to a fast neutron fluence of 10 22 m -2 ( E>0.1 MeV) to investigate the radiation induced degradation of the dielectric strength of the insulation system. After warm-up to room temperature, swelling, weight loss, and the breakdown strength were measured at 77 K. The sandwich swells by 4% at a fluence of 5×10 21 m-2 and by 9% at 1×10 22 m-2. The weight loss of the FRP is 2% at 1×10 22 m-2. The dielectric strength remained unchanged over the whole dose range.

Development of an electrostatic dust detector for tungsten dust

NASA Astrophysics Data System (ADS)

Starkey, D.; Hammond, K.; Roquemore, L.; Skinner, C. H.

2012-10-01

Next-step fusion reactors, such as ITER, are expected to have large quantities of dust that will present hazards that have yet to be encountered in current fusion devices. To manage the amount of dust within the reactors a real-time dust detector must be implemented to ensure that dust does not reach hazardous levels. An electrostatic device that accomplishes this has already been tested on NSTX and Tore Supra [1,2]. We will present modifications of this device to improve its ruggedness to withstand the conditions that will be present in ITER. The detector consists of two tungsten wires wrapped around a macor cylinder that are biased at 100-300 V. Incident dust causes a measurable transient short circuit. Initial results have demonstrated the detection of tungsten particles. We will also present a potential method of electrostatic cleaning of residual dust from the detector.[4pt] [1] C. H. Skinner et al., Rev. Sci. Instrum., 81, 10E102 (2010)[0pt] [2] H. Roche et al., Phys. Scr., T145, (2011).

ETR HEAT EXCHANGER BUILDING, TRA644. FLOOR PLAN AND SECTIONS. PUMP ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. FLOOR PLAN AND SECTIONS. PUMP CUBICLES WITH PUMP MOTORS OUTSIDE CUBICLES. HEAT EXCHANGER EQUIPMENT. COOLANT PIPE TUNNEL ENTERS FROM REACTOR BUILDING. KAISER ETR-5582-MTR-644-A-3, 2/1956. INL INDEX NO. 532-0644-00-486-101294, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. NORTHSOUTH SECTION, LOOKING WEST. STEELFRAME ROOF, CRANE RAIL, ...

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

ETR, TRA-642. NORTH-SOUTH SECTION, LOOKING WEST. STEEL-FRAME ROOF, CRANE RAIL, AND CRANES. COOLANT PIPE TUNNEL LEADING TO REACTOR FROM EAST. (THIS WAS A PRELIMINARY CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-4, 11/1955. INL INDEX NO. 532-0642-00-486-100912, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The Joint European Torus (JET)

NASA Astrophysics Data System (ADS)

Rebut, Paul-Henri

2017-02-01

This paper addresses the history of JET, the Tokamak that reached the highest performances and the experiment that so far came closest to the eventual goal of a fusion reactor. The reader must be warned, however, that this document is not a comprehensive study of controlled thermonuclear fusion or even of JET. The next step on this road, the ITER project, is an experimental reactor. Actually, several prototypes will be required before a commercial reactor can be built. The fusion history is far from been finalised. JET is still in operation some 32 years after the first plasma and still has to provide answers to many questions before ITER takes the lead on research. Some physical interpretations of the observed phenomena, although coherent, are still under discussion. This paper also recalls some basic physics concepts necessary to the understanding of confinement: a knowledgeable reader can ignore these background sections. This fascinating story, comprising successes and failures, is imbedded in the complexities of twentieth and the early twenty-first centuries at a time when world globalization is evolving and the future seems loaded with questions. The views here expressed on plasma confinement are solely those of the author. This is especially the case for magnetic turbulence, for which other scientists may have different views.

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

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

L. C. Cadwallader; C. P. C. Wong; M. Abdou

2014-10-01

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module andmore » blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.« less

Extruder system for high-throughput/steady-state hydrogen ice supply and application for pellet fueling of reactor-scale fusion experiments

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

Combs, S.K.; Foust, C.R.; Qualls, A.L.

Pellet injection systems for the next-generation fusion devices, such as the proposed International Thermonuclear Experimental Reactor (ITER), will require feed systems capable of providing a continuous supply of hydrogen ice at high throughputs. A straightforward concept in which multiple extruder units operate in tandem has been under development at the Oak Ridge National Laboratory. A prototype with three large-volume extruder units has been fabricated and tested in the laboratory. In experiments, it was found that each extruder could provide volumetric ice flow rates of up to {approximately}1.3 cm{sup 3}/s (for {approximately}10 s), which is sufficient for fueling fusion reactors atmore » the gigawatt power level. With the three extruders of the prototype operating in sequence, a steady rate of {approximately}0.33 cm{sup 3}/s was maintained for a duration of 1 h. Even steady-state rates approaching the full ITER design value ({approximately}1 cm{sup 3}/s) may be feasible with the prototype. However, additional extruder units (1{endash}3) would facilitate operations at the higher throughputs and reduce the duty cycle of each unit. The prototype can easily accommodate steady-state pellet fueling of present large tokamaks or other near-term plasma experiments.« less

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

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

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR ...

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

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR STEEL FRAME STRUCTURE. INL NEGATIVE NO. 1330. Unknown Photographer, 1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

DART system analysis.

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

Boggs, Paul T.; Althsuler, Alan; Larzelere, Alex R.

2005-08-01

The Design-through-Analysis Realization Team (DART) is chartered with reducing the time Sandia analysts require to complete the engineering analysis process. The DART system analysis team studied the engineering analysis processes employed by analysts in Centers 9100 and 8700 at Sandia to identify opportunities for reducing overall design-through-analysis process time. The team created and implemented a rigorous analysis methodology based on a generic process flow model parameterized by information obtained from analysts. They also collected data from analysis department managers to quantify the problem type and complexity distribution throughout Sandia's analyst community. They then used this information to develop a communitymore » model, which enables a simple characterization of processes that span the analyst community. The results indicate that equal opportunity for reducing analysis process time is available both by reducing the ''once-through'' time required to complete a process step and by reducing the probability of backward iteration. In addition, reducing the rework fraction (i.e., improving the engineering efficiency of subsequent iterations) offers approximately 40% to 80% of the benefit of reducing the ''once-through'' time or iteration probability, depending upon the process step being considered. Further, the results indicate that geometry manipulation and meshing is the largest portion of an analyst's effort, especially for structural problems, and offers significant opportunity for overall time reduction. Iteration loops initiated late in the process are more costly than others because they increase ''inner loop'' iterations. Identifying and correcting problems as early as possible in the process offers significant opportunity for time savings.« less

CRISPR/Cas9-coupled recombineering for metabolic engineering of Corynebacterium glutamicum.

PubMed

Cho, Jae Sung; Choi, Kyeong Rok; Prabowo, Cindy Pricilia Surya; Shin, Jae Ho; Yang, Dongsoo; Jang, Jaedong; Lee, Sang Yup

2017-07-01

Genome engineering of Corynebacterium glutamicum, an important industrial microorganism for amino acids production, currently relies on random mutagenesis and inefficient double crossover events. Here we report a rapid genome engineering strategy to scarlessly knock out one or more genes in C. glutamicum in sequential and iterative manner. Recombinase RecT is used to incorporate synthetic single-stranded oligodeoxyribonucleotides into the genome and CRISPR/Cas9 to counter-select negative mutants. We completed the system by engineering the respective plasmids harboring CRISPR/Cas9 and RecT for efficient curing such that multiple gene targets can be done iteratively and final strains will be free of plasmids. To demonstrate the system, seven different mutants were constructed within two weeks to study the combinatorial deletion effects of three different genes on the production of γ-aminobutyric acid, an industrially relevant chemical of much interest. This genome engineering strategy will expedite metabolic engineering of C. glutamicum. Copyright © 2017 International Metabolic Engineering Society. Published by Elsevier Inc. All rights reserved.

On-target digestion of collected bacteria for MALDI mass spectrometry.

PubMed

Dugas, Alton J; Murray, Kermit K

2008-10-03

An on-target protein digestion system was developed for the identification of microorganisms in collected bioaerosols using off-line matrix-assisted laser desorption/ionization mass spectrometry (MALDI-MS). Bacteria analysis techniques based on MALDI-MS were adapted for use with an orthogonal MALDI quadrupole-time-of-flight mass spectrometer. Bioaerosols were generated using a pneumatic nebulizer and infused into a chamber for sampling. An Andersen N6 single-stage impactor was used to collect the bioaerosols on a MALDI target. On-target digestion was carried out inside temporary mini-wells placed over the impacted samples. The wells served as miniature reactors for proteolysis. Collected test aerosol particles containing the protein cytochrome c and E. coli bacteria were proteolyzed in situ using trypsin or cyanogen bromide. A total of 19 unique proteins were identified for E. coli. Using the TOF-MS spectra of the digested samples, peptide mass mapping was performed using the MASCOT search engine and an iterative search technique.

Self-consistent modeling of CFETR baseline scenarios for steady-state operation

NASA Astrophysics Data System (ADS)

Chen, Jiale; Jian, Xiang; Chan, Vincent S.; Li, Zeyu; Deng, Zhao; Li, Guoqiang; Guo, Wenfeng; Shi, Nan; Chen, Xi; CFETR Physics Team

2017-07-01

Integrated modeling for core plasma is performed to increase confidence in the proposed baseline scenario in the 0D analysis for the China Fusion Engineering Test Reactor (CFETR). The steady-state scenarios are obtained through the consistent iterative calculation of equilibrium, transport, auxiliary heating and current drives (H&CD). Three combinations of H&CD schemes (NB + EC, NB + EC + LH, and EC + LH) are used to sustain the scenarios with q min > 2 and fusion power of ˜70-150 MW. The predicted power is within the target range for CFETR Phase I, although the confinement based on physics models is lower than that assumed in 0D analysis. Ideal MHD stability analysis shows that the scenarios are stable against n = 1-10 ideal modes, where n is the toroidal mode number. Optimization of RF current drive for the RF-only scenario is also presented. The simulation workflow for core plasma in this work provides a solid basis for a more extensive research and development effort for the physics design of CFETR.

Verification and optimization of the CFETR baseline scenario

NASA Astrophysics Data System (ADS)

Zhao, D.; Lao, L. L.; Meneghini, O.; Staebler, G. M.; Candy, J.; Smith, S. P.; Snyder, P. B.; Prater, R.; Chen, X.; Chan, V. S.; Li, J.; Chen, J.; Shi, N.; Guo, W.; Pan, C.; Jian, X.

2016-10-01

The baseline scenario of China Fusion Engineering Test Reactor (CFETR) was designed starting from 0D calculations. The CFETR baseline scenario satisfies the minimum goal of Fusion Nuclear Science Facility aimed at bridging the gaps between ITER and DEMO. 1.5D calculations are presented to verify the on-going efforts in higher-dimensional modeling of CFETR. Steady-state scenarios are calculated self-consistently by the OMFIT integrated modeling framework that includes EFIT for equilibrium, ONETWO for sources and current, TGYRO for transport. With 68MW of neutral beam power and 8MW of ECH injected to the plasma, the average ion temperature is maintained at 15keV, while 150MW fusion power is produced. The neutral beams also drive 55% of the plasma current. Modest fast ion diffusion will reduce NBCD and affect the profile substantially. Top-launch ECH will increase the current drive and the power absorption rate. EPED model are being included. Work supported by U.S. DOE under DE-FC02-04ER54698 and the USTC CFETR contract.

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.)

Engineering aspects of design and integration of ECE diagnostic in ITER

DOE PAGES

Udintsev, V. S.; Taylor, G.; Pandya, H. K.B.; ...

2015-03-12

ITER ECE diagnostic [1] needs not only to meet measurement requirements, but also to withstand various loads, such as electromagnetic, mechanical, neutronic and thermal, and to be protected from stray ECH radiation at 170 GHz and other millimeter wave emission, like Collective Thomson scattering which is planned to operate at 60 GHz. Same or similar loads will be applied to other millimetre-wave diagnostics [2], located both in-vessel and in-port plugs. These loads must be taken into account throughout the design phases of the ECE and other microwave diagnostics to ensure their structural integrity and maintainability. The integration of microwave diagnosticsmore » with other ITER systems is another challenging activity which is currently ongoing through port integration and in-vessel integration work. Port Integration has to address the maintenance and the safety aspects of diagnostics, too. Engineering solutions which are being developed to support and to operate ITER ECE diagnostic, whilst complying with safety and maintenance requirements, are discussed in this paper.« less

Preliminary study of fusion reactor: Solution of Grad Shapranov equation

NASA Astrophysics Data System (ADS)

Setiawan, Y.; Fermi, N.; Su'ud, Z.

2012-06-01

Nuclear fussion is prospective energy sources for the future due to the abundance of the fuel and can be categorized and clean energy sources. The problem is how to contain very hot plasma of temperature few hundreed million degrees safety and reliably. Tokamax type fussion reactors is considered as the most prospective concept. To analyze the plasma confining process and its movement Grad-Shavranov equation must be solved. This paper discuss about solution of Grad-Shavranov equation using Whittaker function. The formulation is then applied to the ITER design and example.

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

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

Susan Stacy; Hollie K. Gilbert

2005-02-01

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to housemore » the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.« less

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

ETR, TRA642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF ...

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

ETR, TRA-642. CONSOLE FLOOR. CAMERA IS ON WEST SIDE OF FLOOR AND FACES NORTH. OUTER WALL OF STORAGE CANAL IS AT RIGHT. SHIELDING IS THICKER AT LOWER LEVEL, WHERE SPENT FUEL ELEMENTS WILL COOL AFTER REMOVAL FROM REACTOR. INL NEGATIVE NO. 56-1401. Jack L. Anderson, Photographer, 5/1/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CRITICAL FACILITY, TRA654. SCIENTISTS STAND AT EDGE OF TANK ...

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

ETR CRITICAL FACILITY, TRA-654. SCIENTISTS STAND AT EDGE OF TANK AND LIFT REMOVABLE BRIDGE ABOVE THE REACTOR. CONTROL RODS AND FUEL RODS ARE BELOW ENOUGH WATER TO SHIELD WORKERS ABOVE. NOTE CRANE RAILS ALONG WALLS, PUMICE BLOCK WALLS. INL NEGATIVE NO. 57-3690. R.G. Larsen, Photographer, 7/29/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CONTROL BUILDING, TRA647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT ...

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

ETR CONTROL BUILDING, TRA-647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT PANELS AT REAR OF OPERATOR'S CONSOLE GAVE OPERATOR STATUS OF REACTOR PERFORMANCE, COOLANT-WATER CHARACTERISTICS AND OTHER INDICATORS. WINDOWS AT RIGHT LOOKED INTO ETR BUILDING FIRST FLOOR. CAMERA FACING EAST. INL NEGATIVE NO. HD42-6. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ENGINEERING APPLICATIONS OF ANALOG COMPUTERS

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

Bryant, L.T.; Janicke, M.J.; Just, L.C.

1961-02-01

Six examples are given of the application of analog computers in the fields of reactor engineering, heat transfer, and dynamics: deceleration of a reactor control rod by dashpot, pressure variations through a packed bed, reactor kinetics over many decades with thermal feedback (simulation of a TREAT transient), vibrating system with two degrees of freedom, temperature distribution in a radiating fin, and temperature distribution in an irfinite slab with variable thermal properties. (D.L.C.)

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

MTRETR MAINTENANCE SHOP, TRA653. FLOOR PLAN FOR MEZZANINE: LUNCH AND ...

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

MTR-ETR MAINTENANCE SHOP, TRA-653. FLOOR PLAN FOR MEZZANINE: LUNCH AND CONFERENCE ROOM, STORAGE AREA, OFFICES FOR FOREMEN, STENOS, ENGINEERS, DISPATCHER, WOMEN'S RESTROOM. HUMMEL HUMMEL & JONES 810-MTR-ETR-653-A-12, 2/1958. INL INDEX NO. 532-0653-00-381-102837, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NTRE extended life feasibility assessment

NASA Technical Reports Server (NTRS)

1993-01-01

Results of a feasibility analysis of a long life, reusable nuclear thermal rocket engine are presented in text and graph form. Two engine/reactor concepts are addressed: the Particle Bed Reactor (PBR) design and the Commonwealth of Independent States (CIS) concept. Engine design, integration, reliability, and safety are addressed by various members of the NTRE team from Aerojet Propulsion Division, Energopool (Russia), and Babcock & Wilcox.

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...

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

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

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

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG ...

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

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG WALL. WORKERS MAKE ELECTRICAL AND OTHER CONNECTIONS. INL NEGATIVE NO. 4289. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SOUTH WING, TRA661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH ...

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

SOUTH WING, TRA-661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH BAY BEYOND. INL NEGATIVE NO. HD46-45-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. WORKERS ARE INSTALLING HEAT EXCHANGER ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. WORKERS ARE INSTALLING HEAT EXCHANGER PIPING. INL NEGATIVE NO. 56-3122. Jack L. Anderson, Photographer, 9/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING ...

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

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING PLANT. MTR AND ITS ATTACHMENTS IN FOREGROUND. ETR BEYOND TO RIGHT. INL NEGATIVE NO. 56-4100. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR COMPRESSOR BUILDING, TRA643. COMPRESSORS AND OTHER EQUIPMENT INSTALLED. METAL ...

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

ETR COMPRESSOR BUILDING, TRA-643. COMPRESSORS AND OTHER EQUIPMENT INSTALLED. METAL ROOF AND CONCRETE BLOCK WALLS. INL NEGATIVE NO. 61-4536. Unknown Photographer, ca. 1961. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

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

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

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

Ferrada, Juan J; Reiersen, Wayne T

U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C andmore » 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment with no field experience and lowers specific costs while providing higher reliability. This paper presents a brief description of the TCWS conceptual design and the application of RAMI tools to optimize the design at different stages during the project.« less

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

Early Program Development

NASA Image and Video Library

2004-04-15

This artist's concept illustrates the NERVA (Nuclear Engine for Rocket Vehicle Application) engine's hot bleed cycle in which a small amount of hydrogen gas is diverted from the thrust nozzle, thus eliminating the need for a separate system to drive the turbine. The NERVA engine, based on KIWI nuclear reactor technology, would power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which the Marshall Space Flight Center had development responsibility.

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

Methane Dual Expander Aerospike Nozzle Rocket Engine

DTIC Science & Technology

2012-03-22

include O/F ratio, thrust, and engine geometry. After thousands of iterations over the design space , the selected MDEAN engine concept has 349 s of...35 Table 7: Fluid Property Table Supported Parameters...44 Table 8: Fluid Property Input Data Independent Variable Ranges. ................................. 46 Table 9

ETRMTR MECHANICAL SERVICES BUILDING, TRA653. CAMERA FACING NORTHWEST AS BUILDING ...

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

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653. CAMERA FACING NORTHWEST AS BUILDING WAS NEARLY COMPLETE. INL NEGATIVE NO. 57-3653. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Nuclear power in the 21st century: Challenges and possibilities.

PubMed

Horvath, Akos; Rachlew, Elisabeth

2016-01-01

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

Present status of liquid metal research for a fusion reactor

NASA Astrophysics Data System (ADS)

Tabarés, Francisco L.

2016-01-01

Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

NASA Astrophysics Data System (ADS)

Kapychev, V.; Davydov, D.; Gorokhov, V.; Ioltukhovskiy, A.; Kazennov, Yu; Tebus, V.; Frolov, V.; Shikov, A.; Shishkov, N.; Kovalenko, V.; Shishkin, N.; Strebkov, Yu

2000-12-01

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water-graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm2) are considered. At the present time, development and fabrication of lithium orthosilicate-beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350-700°C. Technical problems associated with manufacturing of the modules are discussed.

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

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

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

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

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

PROCESS WATER BUILDING, TRA605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), ...

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

PROCESS WATER BUILDING, TRA-605. FLASH EVAPORATOR, CONDENSER (PROJECT FROM EVAPORATOR), AND STEAM EJECTOR (ALONG REAR WALL). INL NEGATIVE NO. 4377. M.H. Bartz, Photographer, 3/5/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems

EPA Science Inventory

Slide presentation at Conference: ASCE 7th Civil Engineering Conference in the Asian Region. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater & sur...

FAN HOUSE INTERIOR. THREE MOTOR DRIVES FOR POSITIVE DISPLACEMENT BLOWERS ...

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

FAN HOUSE INTERIOR. THREE MOTOR DRIVES FOR POSITIVE DISPLACEMENT BLOWERS LINE UP ON NORTH WALL. CONCRETE PEDESTALS. CAMERA FACES NORTHEAST. INL NEGATIVE NO. 4291. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID ...

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

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID NOT EXPLAIN DARK CLOUD. MTR WING WILL ATTACH TO GROUND FLOOR. INL NEGATIVE NO. 1567. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE SHOWN IN ID-33-G-101. INL NEGATIVE NO. HD24-3-4. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Analysis of Anderson Acceleration on a Simplified Neutronics/Thermal Hydraulics System

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

Toth, Alex; Kelley, C. T.; Slattery, Stuart R

ABSTRACT A standard method for solving coupled multiphysics problems in light water reactors is Picard iteration, which sequentially alternates between solving single physics applications. This solution approach is appealing due to simplicity of implementation and the ability to leverage existing software packages to accurately solve single physics applications. However, there are several drawbacks in the convergence behavior of this method; namely slow convergence and the necessity of heuristically chosen damping factors to achieve convergence in many cases. Anderson acceleration is a method that has been seen to be more robust and fast converging than Picard iteration for many problems, withoutmore » significantly higher cost per iteration or complexity of implementation, though its effectiveness in the context of multiphysics coupling is not well explored. In this work, we develop a one-dimensional model simulating the coupling between the neutron distribution and fuel and coolant properties in a single fuel pin. We show that this model generally captures the convergence issues noted in Picard iterations which couple high-fidelity physics codes. We then use this model to gauge potential improvements with regard to rate of convergence and robustness from utilizing Anderson acceleration as an alternative to Picard iteration.« less

Using the sound of nuclear energy

DOE PAGES

Garrett, Steven; Smith, James; Smith, Robert; ...

2016-08-01

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Using the sound of nuclear energy

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

Garrett, Steven; Smith, James; Smith, Robert

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Nuclear Thermal Propulsion: Past, Present, and a Look Ahead

NASA Technical Reports Server (NTRS)

Borowski, Stanley K.

2014-01-01

NTR: High thrust high specific impulse (2 x LOXLH2 chemical) engine uses high power density fission reactor with enriched uranium fuel as thermal power source. Reactor heat is removed using H2 propellant which is then exhausted to produce thrust. Conventional chemical engine LH2 tanks, turbo pumps, regenerative nozzles and radiation-cooled shirt extensions used -- NTR is next evolutionary step in high performance liquid rocket engines.

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...

11. Photocopy of drawing, February 1958. WATERTOWN ARSENAL REACTOR, SHIELD ...

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

11. Photocopy of drawing, February 1958. WATERTOWN ARSENAL REACTOR, SHIELD STRUCTURE, SECTIONS, AND PLANS. Bendix Aviation Corporation; and Giffels & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan. Drawing Number 53-198. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

ETR ELECTRICAL BUILDING, TRA648. EMERGENCY STANDBY GENERATOR AND DIESEL UNIT. ...

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

ETR ELECTRICAL BUILDING, TRA-648. EMERGENCY STANDBY GENERATOR AND DIESEL UNIT. METAL ROOF AND PUMICE BLOCK WALLS. CAMERA FACING SOUTHWEST. INL NEGATIVE NO. 56-3708. R.G. Larsen, Photographer, 11/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. FLOOR PLANS FOR FIRST FLOOR AND ...

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

ETR ELECTRICAL BUILDING, TRA-648. FLOOR PLANS FOR FIRST FLOOR AND BASEMENT. SECTIONS. KAISER ETR-5528-MTR-648-A-2, 12/1955. INL INDEX NO. 532-0648-00-486-101402, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. CABLE TUNNEL. ...

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

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. CABLE TUNNEL. CAMERA FACING SOUTH INTO ETR ELECTRICAL BUILDING (TRA-648). INL NEGATIVE NO. HD46-20-2. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR ELECTRICAL BUILDING, TRA648. ELEVATIONS AND DETAILS. ROOF PLAN. DOOR ...

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

ETR ELECTRICAL BUILDING, TRA-648. ELEVATIONS AND DETAILS. ROOF PLAN. DOOR SCHEDULE. KAISER ETR-5528-MTR-648-A-3, 1/1956. INL INDEX NO. 532-0648-00-486-101403, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS CAUTION SIGNS BECAUSE OF SODIUM HAZARD WITHIN. INL NEGATIVE NO. HD24-3-1. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING, TRA604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. ...

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

MTR WING, TRA-604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. HVAC AREA IS AT RIGHT OF CORRIDOR. INL NEGATIVE NO. HD46-13-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. DETAIL OF SOUTH SIDE BUILDING ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. DETAIL OF SOUTH SIDE BUILDING INSET. DEMINERALIZER WING AT RIGHT. CAMERA FACING NORTH. INL NEGATIVE NO. HD46-36-2. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PERFORMANCE OF TWO LIQUID METAL TURBOPROP ENGINES UTILIZING A CIRCULATING FUEL REACTOR

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

Tiedemann, H.J.; Mathews, L.

1955-01-20

The performance of two all-nuclear turboprop engines utilizing the circulating fuel reactor with a fluoride fuel temperature of I500 deg F was investigated. Data are presented for off-match-point and modified match-point performances. Results are given in graph form. (M.C.G.)

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

REACTIVITY MEASUREMENT FACILITY, UNDER CONSTRUCTION OVER MTR CANAL IN BASEMENT ...

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

REACTIVITY MEASUREMENT FACILITY, UNDER CONSTRUCTION OVER MTR CANAL IN BASEMENT OF MTR BUILDING, TRA-603. WOOD PLANKS REST ON CANAL WALL OBSERVABLE IN FOREGROUND. INL NEGATIVE NO. 11745. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING, TRA604. PRECAST CONCRETE PANELS AND DIMENSIONS FOR PANELS ...

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

MTR WING, TRA-604. PRECAST CONCRETE PANELS AND DIMENSIONS FOR PANELS K THROUGH Q. BLAW-KNOX 3150-804-21, SHEET #2, 11/1950. INL INDEX NO. 531-0604-62-098-100645, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NORTH BASEMENT WALL. IBEAM COLUMNS HAVE BEEN ENCASED IN CONCRETE. ...

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

NORTH BASEMENT WALL. I-BEAM COLUMNS HAVE BEEN ENCASED IN CONCRETE. STEEL BEAMS LAY ACROSS FIRST FLOOR AWAITING CONCRETE POUR. CAMERA LOOKS SOUTHWEST. INL NEGATIVE NO. 735. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. ONE OF THREE EVAPORATORS BEFORE IT ...

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

PROCESS WATER BUILDING, TRA-605. ONE OF THREE EVAPORATORS BEFORE IT IS INSTALLED IN UPPER LEVEL OF EAST HALF OF BUILDING. INL NEGATIVE NO. 1533. Unknown Photographer, 3/1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. BENCH MARK AND ELEVATION LOCATIONS, FLOOR LOADING DATA, ...

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

ETR, TRA-642. BENCH MARK AND ELEVATION LOCATIONS, FLOOR LOADING DATA, CRANE WORKING AREAS. PHILLIPS PETROLEUM COMPANY ETR-D-1584, 5/1959. INL INDEX NO. 532-0642-00-706-020323, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

NASA Astrophysics Data System (ADS)

Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

2018-06-01

The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

Applying design principles to fusion reactor configurations for propulsion in space

NASA Technical Reports Server (NTRS)

Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

1993-01-01

The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

The novel implicit LU-SGS parallel iterative method based on the diffusion equation of a nuclear reactor on a GPU cluster

NASA Astrophysics Data System (ADS)

Zhang, Jilin; Sha, Chaoqun; Wu, Yusen; Wan, Jian; Zhou, Li; Ren, Yongjian; Si, Huayou; Yin, Yuyu; Jing, Ya

2017-02-01

GPU not only is used in the field of graphic technology but also has been widely used in areas needing a large number of numerical calculations. In the energy industry, because of low carbon, high energy density, high duration and other characteristics, the development of nuclear energy cannot easily be replaced by other energy sources. Management of core fuel is one of the major areas of concern in a nuclear power plant, and it is directly related to the economic benefits and cost of nuclear power. The large-scale reactor core expansion equation is large and complicated, so the calculation of the diffusion equation is crucial in the core fuel management process. In this paper, we use CUDA programming technology on a GPU cluster to run the LU-SGS parallel iterative calculation against the background of the diffusion equation of the reactor. We divide one-dimensional and two-dimensional mesh into a plurality of domains, with each domain evenly distributed on the GPU blocks. A parallel collision scheme is put forward that defines the virtual boundary of the grid exchange information and data transmission by non-stop collision. Compared with the serial program, the experiment shows that GPU greatly improves the efficiency of program execution and verifies that GPU is playing a much more important role in the field of numerical calculations.

Individual dose due to radioactivity accidental release from fusion reactor.

PubMed

Nie, Baojie; Ni, Muyi; Wei, Shiping

2017-04-05

As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1kg HTO and 1000kg dust release) and 34-52years for case 2 (1kg HTO and 10kg-100kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation. Copyright © 2016 Elsevier B.V. All rights reserved.

Conceptual design of ACB-CP for ITER cryogenic system

NASA Astrophysics Data System (ADS)

Jiang, Yongcheng; Xiong, Lianyou; Peng, Nan; Tang, Jiancheng; Liu, Liqiang; Zhang, Liang

2012-06-01

ACB-CP (Auxiliary Cold Box for Cryopumps) is used to supply the cryopumps system with necessary cryogen in ITER (International Thermonuclear Experimental Reactor) cryogenic distribution system. The conceptual design of ACB-CP contains thermo-hydraulic analysis, 3D structure design and strength checking. Through the thermohydraulic analysis, the main specifications of process valves, pressure safety valves, pipes, heat exchangers can be decided. During the 3D structure design process, vacuum requirement, adiabatic requirement, assembly constraints and maintenance requirement have been considered to arrange the pipes, valves and other components. The strength checking has been performed to crosscheck if the 3D design meets the strength requirements for the ACB-CP.

Learning to Teach Elementary Science through Iterative Cycles of Enactment in Culturally and Linguistically Diverse Contexts

ERIC Educational Resources Information Center

Bottoms, SueAnn I.; Ciechanowski, Kathryn M.; Hartman, Brian

2015-01-01

Iterative cycles of enactment embedded in culturally and linguistically diverse contexts provide rich opportunities for preservice teachers (PSTs) to enact core practices of science. This study is situated in the larger Families Involved in Sociocultural Teaching and Science, Technology, Engineering and Mathematics (FIESTAS) project, which weaves…

VERA 3.6 Release Notes

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

Williamson, Richard L.; Kochunas, Brendan; Adams, Brian M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms.

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA603. SUMMARY OF ...

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

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA-603. SUMMARY OF COOLANT FLOW FROM WORKING RESERVOIR TO INTERIOR OF REACTOR'S THERMAL SHIELD. NAMES TANK SECTIONS. PIPE AND DRAIN-LINE SIZES. SHOWS DIRECTION OF AIR FLOW THROUGH PEBBLE AND GRAPHITE BLOCK ZONE. NEUTRON CURTAIN AND THERMAL COLUMN DOOR. BLAW-KNOX 3150-92-7, 3/1950. INL INDEX NO. 531-0603-51-098-100036, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

NASA Technical Reports Server (NTRS)

Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

1973-01-01

A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

A Generalized Perturbation Theory Solver In Rattlesnake Based On PETSc With Application To TREAT Steady State Uncertainty Quantification

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

Schunert, Sebastian; Wang, Congjian; Wang, Yaqi

Rattlesnake and MAMMOTH are the designated TREAT analysis tools currently being developed at the Idaho National Laboratory. Concurrent with development of the multi-physics, multi-scale capabilities, sensitivity analysis and uncertainty quantification (SA/UQ) capabilities are required for predicitive modeling of the TREAT reactor. For steady-state SA/UQ, that is essential for setting initial conditions for the transients, generalized perturbation theory (GPT) will be used. This work describes the implementation of a PETSc based solver for the generalized adjoint equations that constitute a inhomogeneous, rank deficient problem. The standard approach is to use an outer iteration strategy with repeated removal of the fundamental modemore » contamination. The described GPT algorithm directly solves the GPT equations without the need of an outer iteration procedure by using Krylov subspaces that are orthogonal to the operator’s nullspace. Three test problems are solved and provide sufficient verification for the Rattlesnake’s GPT capability. We conclude with a preliminary example evaluating the impact of the Boron distribution in the TREAT reactor using perturbation theory.« less

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Measurement of the complex transmittance of large optical elements with Ptychographical Iterative Engine.

PubMed

Wang, Hai-Yan; Liu, Cheng; Veetil, Suhas P; Pan, Xing-Chen; Zhu, Jian-Qiang

2014-01-27

Wavefront control is a significant parameter in inertial confinement fusion (ICF). The complex transmittance of large optical elements which are often used in ICF is obtained by computing the phase difference of the illuminating and transmitting fields using Ptychographical Iterative Engine (PIE). This can accurately and effectively measure the transmittance of large optical elements with irregular surface profiles, which are otherwise not measurable using commonly used interferometric techniques due to a lack of standard reference plate. Experiments are done with a Continue Phase Plate (CPP) to illustrate the feasibility of this method.

MTR BUILDING, TRA603. SOUTHEAST CORNER, EAST SIDE FACING TOWARD RIGHT ...

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

MTR BUILDING, TRA-603. SOUTHEAST CORNER, EAST SIDE FACING TOWARD RIGHT OF VIEW. CAMERA FACING NORTHWEST. LIGHT-COLORED PROJECTION AT LEFT IS ENGINEERING SERVICES BUILDING, TRA-635. SMALL CONCRETE BLOCK BUILDING AT CENTER OF VIEW IS FAST CHOPPER DETECTOR HOUSE, TRA-665. INL NEGATIVE NO. HD46-43-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

NASA Technical Reports Server (NTRS)

Clark, John S.; Walton, James T.; Mcguire, Melissa L.

1992-01-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Breadboard RL10-2B low-thrust operating mode (second iteration) test report

NASA Technical Reports Server (NTRS)

Kanic, Paul G.; Kaldor, Raymond B.; Watkins, Pia M.

1988-01-01

Cryogenic rocket engines requiring a cooling process to thermally condition the engine to operating temperature can be made more efficient if cooling propellants can be burned. Tank head idle and pumped idle modes can be used to burn propellants employed for cooling, thereby providing useful thrust. Such idle modes required the use of a heat exchanger to vaporize oxygen prior to injection into the combustion chamber. During December 1988, Pratt and Whitney conducted a series of engine hot firing demonstrating the operation of two new, previously untested oxidizer heat exchanger designs. The program was a second iteration of previous low thrust testing conducted in 1984, during which a first-generation heat exchanger design was used. Although operation was demonstrated at tank head idle and pumped idle, the engine experienced instability when propellants could not be supplied to the heat exchanger at design conditions.

Beryllium for fusion application - recent results

NASA Astrophysics Data System (ADS)

Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

2002-12-01

The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

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

Ghoos, K., E-mail: kristel.ghoos@kuleuven.be; Dekeyser, W.; Samaey, G.

2016-10-01

The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracymore » by making use of averaging in the Random Noise coupling technique.« less

Influence of the optical fiber type on the performances of fiber-optics current sensor dedicated to plasma current measurement in ITER.

PubMed

Aerssens, Matthieu; Descamps, Frédéric; Gusarov, Andrei; Mégret, Patrice; Moreau, Philippe; Wuilpart, Marc

2015-07-01

In this paper, we compare, by means of simulations using the Jones formalism, the performances of several optical fiber types (low birefringence and spun fibers) for the measurement of plasma current in international thermonuclear experimental reactor (ITER). The main results presented in this paper concern the minimum value of the ratio between the beat length and the spun period, which allows meeting the ITER current measurement specifications. Assuming a high-birefringence spun fiber with a beat length of 3 mm, we demonstrate that the minimum ratio between the beat length and the spun period is 4.4 when considering a 28 m long sensing fiber surrounding the vacuum vessel. This minimum ratio rises to 10.14 when a 100 m long lead fiber connecting the interrogating system to the sensing fiber is taken into account.

Assessment of possible failure modes and non-destructive examination of the ITER pre-compression rings

NASA Astrophysics Data System (ADS)

Knaster, J.; Evans, D.; Rajainmaki, H.

2012-06-01

The pre-compression rings (PCRs) for the International Thermonuclear Experimental Reactor (ITER) represent one of the largest and most highly stressed composite structures ever designed for long term operation at 4K. Three rings, each 5m in diameter and 337 × 288 mm in cross-section, will be installed at the top and bottom of the eighteen "D" shaped Toroidal Field (TF) coils to apply a total centripetal load of 70 MN per TF coil. The interaction of the 68 kA conductor current circulating in the coil (for a total of 9.1MA) with the required magnetic field to confine the plasma during operation will result in Lorentz forces that build in-plane and out-of-plane loads. The PCRs are essential to keep the stresses below the acceptable level for the ITER magnets structural materials.

CONTROL OF VOLATILE ORGANIC COMPOUNDS BY AN AC ENERGIZED FERROELECTRIC PELLET REACTOR AND A PULSED CORONA REACTOR

EPA Science Inventory

The paper gives results of a study to develop baseline engineering data to demonstrate the feasibility of application of plasma reactors to the destruction of various volatile organic compounds at ppm levels. Two laboratory-scale reactors, an alternating current energized ferroel...

157. ARAIII Reactor building (ARA608) Main gas loop mechanical flow ...

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

157. ARA-III Reactor building (ARA-608) Main gas loop mechanical flow sheet. This drawing was selected as a typical example of mechanical arrangements within reactor building. Aerojet-general 880-area/GCRE-0608-50-013-102634. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

CUTS FOR MTR EXCAVATION ILLUSTRATE SEDIMENTARY MANTLE OF SOIL AND ...

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

CUTS FOR MTR EXCAVATION ILLUSTRATE SEDIMENTARY MANTLE OF SOIL AND GRAVEL OVERLAYING LAVA ROCK FIFTY FEET BELOW. SAGEBRUSH HAS BEEN SCOURED FROM REST OF SITE. CAMERA PROBABLY FACES SOUTHWEST. INL NEGATIVE NO. 67. Unknown Photographer, 6/4/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605, INTERIOR. FIRST FLOOR. ELECTRICAL EQUIPMENT IN ...

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

PROCESS WATER BUILDING, TRA-605, INTERIOR. FIRST FLOOR. ELECTRICAL EQUIPMENT IN LEFT HALF OF VIEW. CAMERA IS IN NORTHWEST CORNER FACING SOUTHEAST. INL NEGATIVE NO. HD46-27-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. FIRST FLOOR. INSIDE UTILITY CORRIDOR ALONG ...

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

ETR BUILDING, TRA-642, INTERIOR. FIRST FLOOR. INSIDE UTILITY CORRIDOR ALONG SOUTH PERIMETER WALL (COMMON TO ELECTRICAL BUILDING, TRA-648). CAMERA FACES WEST. INL NEGATIVE NO. HD46-16-2. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ELEVATIONS. METAL SIDING. OFFICE BUILDING (TRA647) AND ELECTRICAL ...

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

ETR, TRA-642. ELEVATIONS. METAL SIDING. OFFICE BUILDING (TRA-647) AND ELECTRICAL BUILDING (TRA-648) ATTACHED. KAISER ETR-5528-MTR-642-A-11, 11/1955. INL INDEX NO. 532-0642-00-486-100919, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. WALL SECTION DETAILS. METAL SIDING JOINS TO ELECTRICAL ...

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

ETR, TRA-642. WALL SECTION DETAILS. METAL SIDING JOINS TO ELECTRICAL BUILDING, OFFICE BUILDING, AND ROOF. KAISER ETR-5528-MTR-A-13, 11/1955. INL INDEX NO. 532-0642-00-486-100920, REV. 4. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. FLOOR PLAN UNDER BALCONY ON CONSOLE FLOOR. MOTORGENERATOR ...

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

ETR, TRA-642. FLOOR PLAN UNDER BALCONY ON CONSOLE FLOOR. MOTOR-GENERATOR SETS AND OTHER ELECTRICAL EQUIPMENT. PHILLIPS PETROLEUM COMPANY ETR-D-1781, 7/1960. INL INDEX NO. 532-0642-00-706-020384, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

SOUTH WING, MTR661. INTERIOR DETAIL INSIDE LAB ROOM 131. CAMERA ...

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

SOUTH WING, MTR-661. INTERIOR DETAIL INSIDE LAB ROOM 131. CAMERA FACING NORTHEAST. NOTE CONCRETE BLOCK WALLS. SAFETY SHOWER AND EYE WASHER AT REAR WALL. INL NEGATIVE NO. HD46-7-2. Mike Crane, Photographer, 2/2005. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HOT CELL BUILDING, TRA632. ELEVATIONS FOR SOUTH, NORTH AND WEST ...

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

HOT CELL BUILDING, TRA-632. ELEVATIONS FOR SOUTH, NORTH AND WEST SIDES OF 1958 EXTENSION. H.K. FERGUSON CO. 895-MTR-ETR-632-A3, 12/1958. INL INDEX NO. 531-0632-00-279-101926, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

CANAL EMERGES FROM EAST SIDE OF MTR BUILDING. "EXTRA" LENGTH ...

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

CANAL EMERGES FROM EAST SIDE OF MTR BUILDING. "EXTRA" LENGTH WAS TO STORE SPENT FUEL THAT WOULD ACCUMULATE BEFORE THE CHEMICAL PROCESSING PLANT WAS READY TO PROCESS IT. INL NEGATIVE NO. 1659. Unknown Photographer, 3/9/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

STEEL BEAMS FOR FIRST FLOOR BEING READIED FOR CONCRETE POUR ...

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

STEEL BEAMS FOR FIRST FLOOR BEING READIED FOR CONCRETE POUR UNDER WEATHER SHELTER DURING COLD WINTER. NOTE ABUNDANCE OF BEAMS; THE FLOOR WILL SUPPORT HEAVY LOADS. INL NEGATIVE NO. 1175. Unknown Photographer, 12/20/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CRITICAL FACILITY (ETRCF), TRA654. SOUTH SIDE. CAMERA FACING NORTH ...

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

ETR CRITICAL FACILITY (ETR-CF), TRA-654. SOUTH SIDE. CAMERA FACING NORTH AND ROLL-UP DOOR. ORIGINAL SIDING HAS BEEN REPLACED WITH STUCCO-LIKE MATERIAL. INL NEGATIVE NO. HD46-40-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING A, TRA604, INTERIOR. MAIN FLOOR. DETAIL VIEW INSIDE ...

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

MTR WING A, TRA-604, INTERIOR. MAIN FLOOR. DETAIL VIEW INSIDE LABORATORY 114. CAMERA FACING NORTH. DISPOSAL OF RADIOACTIVE MATERIALS IS UNDERWAY. INL NEGATIVE NO. HD46-12-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

COMPRESSOR BUILDING, TRA626. ELEVATIONS. WINDOWS. WALL SECTIONS. PUMICE BLOCK BUILDING ...

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

COMPRESSOR BUILDING, TRA-626. ELEVATIONS. WINDOWS. WALL SECTIONS. PUMICE BLOCK BUILDING HOUSED COMPRESSORS FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENTS. MTR-626-IDO-2S, 3/1952. INL INDEX NO. 531-0626-00-396-110535, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HOT CELL BUILDING, TRA632, INTERIOR. CELL 3, "HEAVY" CELL. CAMERA ...

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

HOT CELL BUILDING, TRA-632, INTERIOR. CELL 3, "HEAVY" CELL. CAMERA FACES WEST TOWARD BUILDING EXIT. OBSERVATION WINDOW AT LEFT EDGE OF VIEW. INL NEGATIVE NO. HD46-28-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, 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

STORAGE AND RECIEVING, TRA662. ELEVATIONS. LOWBAY SECTION ON SOUTH SIDE ...

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

STORAGE AND RECIEVING, TRA-662. ELEVATIONS. LOW-BAY SECTION ON SOUTH SIDE WAS FLAMMABLE STORAGE AREA. HUMMEL HUMMEL & JONES 1038-MTR-ETR-662-A-3, 6/1960. INL INDEX NO. 532-0653-00-381-102036, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Airbreathing engine selection criteria for SSTO propulsion system

NASA Astrophysics Data System (ADS)

Ohkami, Yoshiaki; Maita, Masataka

1995-02-01

This paper presents airbreathing engine selection criteria to be applied to the propulsion system of a Single Stage To Orbit (SSTO). To establish the criteria, a relation among three major parameters, i.e., delta-V capability, weight penalty, and effective specific impulse of the engine subsystem, is derived as compared to these parameters of the LH2/LOX rocket engine. The effective specific impulse is a function of the engine I(sub sp) and vehicle thrust-to-drag ratio which is approximated by a function of the vehicle velocity. The weight penalty includes the engine dry weight, cooling subsystem weight. The delta-V capability is defined by the velocity region starting from the minimum operating velocity up to the maximum velocity. The vehicle feasibility is investigated in terms of the structural and propellant weights, which requires an iteration process adjusting the system parameters. The system parameters are computed by iteration based on the Newton-Raphson method. It has been concluded that performance in the higher velocity region is extremely important so that the airbreathing engines are required to operate beyond the velocity equivalent to the rocket engine exhaust velocity (approximately 4500 m/s).

52. ARAII. Support piers for SL1 reactor building. September 5, ...

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

52. ARA-II. Support piers for SL-1 reactor building. September 5, 1957. Ineel photo no. 57-4398. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

NASA Astrophysics Data System (ADS)

Deepak, SHARMA; Paritosh, CHAUDHURI

2018-04-01

The Indian test blanket module (TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the R&D activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices (ITER relevant and DEMO). The Indian Lead–Lithium Cooled Ceramic Breeder (LLCB) blanket concept is one of the Indian DEMO relevant TBM, to be tested in ITER as a part of the TBM program. Helium-Cooled Ceramic Breeder (HCCB) is an alternative blanket concept that consists of lithium titanate (Li2TiO3) as ceramic breeder (CB) material in the form of packed pebble beds and beryllium as the neutron multiplier. Specifically, attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions. These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.

Method and apparatus for monitoring a hydrocarbon-selective catalytic reduction device

DOEpatents

Schmieg, Steven J; Viola, Michael B; Cheng, Shi-Wai S; Mulawa, Patricia A; Hilden, David L; Sloane, Thompson M; Lee, Jong H

2014-05-06

A method for monitoring a hydrocarbon-selective catalytic reactor device of an exhaust aftertreatment system of an internal combustion engine operating lean of stoichiometry includes injecting a reductant into an exhaust gas feedstream upstream of the hydrocarbon-selective catalytic reactor device at a predetermined mass flowrate of the reductant, and determining a space velocity associated with a predetermined forward portion of the hydrocarbon-selective catalytic reactor device. When the space velocity exceeds a predetermined threshold space velocity, a temperature differential across the predetermined forward portion of the hydrocarbon-selective catalytic reactor device is determined, and a threshold temperature as a function of the space velocity and the mass flowrate of the reductant is determined. If the temperature differential across the predetermined forward portion of the hydrocarbon-selective catalytic reactor device is below the threshold temperature, operation of the engine is controlled to regenerate the hydrocarbon-selective catalytic reactor device.

Experimentation in software engineering

NASA Technical Reports Server (NTRS)

Basili, V. R.; Selby, R. W.; Hutchens, D. H.

1986-01-01

Experimentation in software engineering supports the advancement of the field through an iterative learning process. In this paper, a framework for analyzing most of the experimental work performed in software engineering over the past several years is presented. A variety of experiments in the framework is described and their contribution to the software engineering discipline is discussed. Some useful recommendations for the application of the experimental process in software engineering are included.

Design of a Rail Gun System for Mitigating Disruptions in Fusion Reactors

NASA Astrophysics Data System (ADS)

Lay, Wei-Siang

Magnetic fusion devices, such as the tokamak, that carry a large amount of current to generate the plasma confining magnetic fields have the potential to lose magnetic stability control. This can lead to a major plasma disruption, which can cause most of the stored plasma energy to be lost to localized regions on the walls, causing severe damage. This is the most important issue for the $20B ITER device (International Thermonuclear Experimental Reactor) that is under construction in France. By injecting radiative materials deep into the plasma, the plasma energy could be dispersed more evenly on the vessel surface thus mitigating the harmful consequences of a disruption. Methods currently planned for ITER rely on the slow expansion of gases to propel the radiative payloads, and they also need to be located far away from the reactor vessel, which further slows down the response time of the system. Rail guns are being developed for aerospace applications, such as for mass transfer from the surface of the moon and asteroids to low earth orbit. A miniatured version of this aerospace technology seems to be particularly well suited to meet the fast time response needs of an ITER disruption mitigation system. Mounting this device close to the reactor vessel is also possible, which substantially increases its performance because the stray magnetic fields near the vessel walls could be used to augment the rail gun generated magnetic fields. In this thesis, the potential viability on Rail Gun based DMS is studied to investigate its projected fast time response capability by design, fabrication, and experiment of an NSTX-U sized rail gun system. Material and geometry based tests are used to find the most suitable armature design for this system for which the desirable attributes are high specific stiffness and high electrical conductivity. With the best material in these studies being aluminum 7075, the experimental Electromagnetic Particle Injector (EPI) system has propelled an aluminum armature (weighing 3g) to a velocity more than 150 m/s within two milliseconds post trigger, consistent with the predicted projection for a system with those parameters. Fixed magnetic field probes and high-speed images capture the velocity profile. To propel the armatures, a 20 mF capacitor bank charged to 2 kV and augmented with external field coils powers the rails. These studies indicate that an EPI based system can indeed operate with a fast response time of less than three milliseconds after an impending disruption is detected, and thus warrants further studies to more fully develop the concept as a back-up option for an ITER DMS.

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

NASA Astrophysics Data System (ADS)

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

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

Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited)a)

NASA Astrophysics Data System (ADS)

Smith, Roger J.

2008-10-01

A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local Bpol diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local Te, ne, and B∥ along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher neB∥ product and higher ne and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

Impurity re-distribution in the corner regions of the JET divertor

NASA Astrophysics Data System (ADS)

Widdowson, A.; Coad, J. P.; Alves, E.; Baron-Wiechec, A.; Barradas, N. P.; Catarino, N.; Corregidor, V.; Heinola, K.; Krat, S.; Likonen, J.; Matthews, G. F.; Mayer, M.; Petersson, P.; Rubel, M.; Contributors, JET

2017-12-01

The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)—the JET ITER-like wall (ILW)—the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

Plasma cleaning of ITER first mirrors

NASA Astrophysics Data System (ADS)

Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

2017-12-01

Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

Optimization applications in aircraft engine design and test

NASA Technical Reports Server (NTRS)

Pratt, T. K.

1984-01-01

Starting with the NASA-sponsored STAEBL program, optimization methods based primarily upon the versatile program COPES/CONMIN were introduced over the past few years to a broad spectrum of engineering problems in structural optimization, engine design, engine test, and more recently, manufacturing processes. By automating design and testing processes, many repetitive and costly trade-off studies have been replaced by optimization procedures. Rather than taking engineers and designers out of the loop, optimization has, in fact, put them more in control by providing sophisticated search techniques. The ultimate decision whether to accept or reject an optimal feasible design still rests with the analyst. Feedback obtained from this decision process has been invaluable since it can be incorporated into the optimization procedure to make it more intelligent. On several occasions, optimization procedures have produced novel designs, such as the nonsymmetric placement of rotor case stiffener rings, not anticipated by engineering designers. In another case, a particularly difficult resonance contraint could not be satisfied using hand iterations for a compressor blade, when the STAEBL program was applied to the problem, a feasible solution was obtained in just two iterations.

Incorporating Prototyping and Iteration into Intervention Development: A Case Study of a Dining Hall-Based Intervention

ERIC Educational Resources Information Center

McClain, Arianna D.; Hekler, Eric B.; Gardner, Christopher D.

2013-01-01

Background: Previous research from the fields of computer science and engineering highlight the importance of an iterative design process (IDP) to create more creative and effective solutions. Objective: This study describes IDP as a new method for developing health behavior interventions and evaluates the effectiveness of a dining hall--based…

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

156. ARAIII Reactor building (ARA608) Electrical and control details of ...

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

156. ARA-III Reactor building (ARA-608) Electrical and control details of mobile work bridge over reactor and pipiing pits. Aerojet-general 880-area/GCRE-608-E-6. Date: November 1958. Ineel index code no. 063-0608-10-013-102621. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

PBF Reactor Building (PER620). Reactor vessel arrives from gate city ...

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

PBF Reactor Building (PER-620). Reactor vessel arrives from gate city steel at door of PBF. On flatbed, it is too high to fit under door. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-737 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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.

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILELIKE STEPS) PROJECTS ...

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

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILE-LIKE STEPS) PROJECTS FROM THE SOUTHEAST CORNER OF THE MTR TOWARD SOUTHEAST CORNER OF BUILDING, WHERE SHIELDING BLOCKS BEGIN TO SURROUND THE TUNNEL AS IT NEARS DETECTING INSTRUMENTS NEAR THE BUILDING WALL. GEAR RELATED TO CRYSTAL NEUTRON SPECTROMETER IS IN FOREGROUND SURROUNDED BY SHIELDING. DATA CONSOLES ARE AT MID-LEVEL OF EAST FACE. OTHER WORK PROCEEDS ON TOP OF AND ELSEWHERE AROUND REACTOR. NOTE TOOLS HANGING AGAINST SOUTHEAST CORNER, USED TO CHANGE FUEL ELEMENTS AND OTHER REACTOR ITEMS DURING REFUELING CYCLES. INL NEGATIVE NO. 10439. Unknown Photographer, 4/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. CORRIDOR ALONG WEST WALL OF ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. CORRIDOR ALONG WEST WALL OF BUILDING, WHICH IS AT RIGHT OF VIEW. AUDIO ALARM IS ALONG WALL AT RIGHT. CAMERA FACES SOUTH. INL NEGATIVE NO. HD46-30-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING A, TRA604. SOUTH SIDE. CAMERA FACING NORTH. THIS ...

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

MTR WING A, TRA-604. SOUTH SIDE. CAMERA FACING NORTH. THIS VIEW TYPIFIES TENDENCY FOR EXPANSIONS TO TAKE THE FORM OF PROJECTIONS AND INFILL USING AVAILABLE YARD SPACES. INL NEGATIVE NO. HD47-44-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR, TRA603. CONTROL ROOM DETAILS. ACOUSTIC PLASTER CEILING, USHAPED CONSOLE, ...

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

MTR, TRA-603. CONTROL ROOM DETAILS. ACOUSTIC PLASTER CEILING, U-SHAPED CONSOLE, INSTRUMENT PANELS, GLASS DOOR, ASPHALT TILE FLOOR AND COLORS. BLAW-KNOX 3150-803-11, 10/1950. INL INDEX NO. 531-0603-00-098-100570, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HEAT EXCHANGER BUILDING, TRA644. NORTHEAST CORNER. CAMERA IS ON PIKE ...

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

HEAT EXCHANGER BUILDING, TRA-644. NORTHEAST CORNER. CAMERA IS ON PIKE STREET FACING SOUTHWEST. ATTACHED STRUCTURE AT RIGHT OF VIEW IS ETR COMPRESSOR BUILDING, TRA-643. INL NEGATIVE NO. HD46-36-4. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETRMTR MECHANICAL SERVICES BUILDING, TRA653, INTERIOR. CAMERA IS INSIDE MEN'S ...

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

ETR-MTR MECHANICAL SERVICES BUILDING, TRA-653, INTERIOR. CAMERA IS INSIDE MEN'S LAVATORY AND SHOWER FACING SOUTHEAST. SHOWER AND TOILET STALLS ARE IN PLACE. ROUND COMMUNAL SINK AT LEFT. INL NEGATIVE NO. 57-3652. K. Mansfield, Photographer, 7/22/1957 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING, TRA604. BASEMENT FLOOR PLAN. FIREPROOF RECORD ROOM BELOW ...

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

MTR WING, TRA-604. BASEMENT FLOOR PLAN. FIRE-PROOF RECORD ROOM BELOW COUNTING ROOM. HEATING AND COOLING EQUIPMENT. UNSPECIFIED EXPANSION AREA ALONG WEST WALL. BLAW-KNOX 3150-4-1, 7/1950. INL INDEX NO. 531-0604-00-098-100007, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. CAMERA LOOKING EAST AND TO WEST ...

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

PROCESS WATER BUILDING, TRA-605. CAMERA LOOKING EAST AND TO WEST WALL NOW ENCLOSING FLASH EVAPORATORS. PIPES IN FOREGROUND WILL CARRY DEMINERALIZED COOLING WATER TO AND FROM THE MTR. INL NEGATIVE NO. 2937. Unknown Photographer, 7/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

HORIZONTAL BEAM HOLE NO. 3. PLUG AND RADIATION DOOR HAVE ...

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

HORIZONTAL BEAM HOLE NO. 3. PLUG AND RADIATION DOOR HAVE BEEN REMOVED. EXPERIMENTAL APPARATUS WAS INSERTED INTO THE HOLE. NOTE VALVE CUBICLES NEAR FLOOR ON EACH SIDE OF HB-3. INL NEGATIVE NO. 3471. Unknown Photographer, 10/12/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Engineering of solar photocatalytic detoxification and disinfection process

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

Goswami, D.Y.

1995-12-31

Use of solar radiation for photocatalytic detoxification and disinfection is a very fascinating and fast-developing area. Although scientific research on these processes, especially photocatalytic oxidation, has been conducted for at least the last three decades, the development of industrial/commercial applications, engineering systems and engineering design methodologies have occurred only recently. A number of reactor concepts and designs, including concentrating and non-concentrating types and various methods of catalyst deployment have been developed. Some of these reactors have been used in field demonstrations of groundwater and wastewater remediation. Recent research has been focused on improvements of catalysts to increase the reaction rates,more » as well as finding new applications of the process. This paper reviews the latest developments of solar detoxification and disinfection including catalyst development, industrial/commercial applications, reactor design and engineering system design methodologies. 80 refs., 20 figs., 3 tabs.« less

The Iterative Design Process in Research and Development: A Work Experience Paper

NASA Technical Reports Server (NTRS)

Sullivan, George F. III

2013-01-01

The iterative design process is one of many strategies used in new product development. Top-down development strategies, like waterfall development, place a heavy emphasis on planning and simulation. The iterative process, on the other hand, is better suited to the management of small to medium scale projects. Over the past four months, I have worked with engineers at Johnson Space Center on a multitude of electronics projects. By describing the work I have done these last few months, analyzing the factors that have driven design decisions, and examining the testing and verification process, I will demonstrate that iterative design is the obvious choice for research and development projects.

PROCESS WATER BUILDING, TRA605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES ...

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

PROCESS WATER BUILDING, TRA-605. AERIAL TAKEN WHILE SEVERAL PIPE TRENCHES REMAINED OPEN. CAMERA FACES EASTERLY. NOTE DUAL PIPES BETWEEN REACTOR BUILDING AND NORTH SIDE OF PROCESS WATER BUILDING. PIPING NEAR WORKING RESERVOIR HEADS FOR RETENTION RESERVOIR. PIPE FROM DEMINERALIZER ENTERS MTR FROM NORTH. SEE ALSO TRENCH FOR COOLANT AIR DUCT AT SOUTH SIDE OF MTR AND LEADING TO FAN HOUSE AND STACK. INL NEGATIVE NO. 2966-A. Unknown Photographer, 7/31/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR CRITICAL FACILITY, TRA654. CONTEXTUAL VIEW. CAMERA ON ROOF OF ...

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

ETR CRITICAL FACILITY, TRA-654. CONTEXTUAL VIEW. CAMERA ON ROOF OF MTR BUILDING AND FACING SOUTH. ETR AND ITS COOLANT BUILDING AT UPPER PART OF VIEW. ETR COOLING TOWER NEAR TOP EDGE OF VIEW. EXCAVATION AT CENTER IS FOR ETR CF. CENTER OF WHICH WILL CONTAIN POOL FOR REACTOR. NOTE CHOPPER TUBE PROCEEDING FROM MTR IN LOWER LEFT OF VIEW, DIAGONAL TOWARD LEFT. INL NEGATIVE NO. 56-4227. Jack L. Anderson, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Status and problems of fusion reactor development.

PubMed

Schumacher, U

2001-03-01

Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

Genome-Resolved Meta-Omics Ties Microbial Dynamics to Process Performance in Biotechnology for Thiocyanate Degradation.

PubMed

Kantor, Rose S; Huddy, Robert J; Iyer, Ramsunder; Thomas, Brian C; Brown, Christopher T; Anantharaman, Karthik; Tringe, Susannah; Hettich, Robert L; Harrison, Susan T L; Banfield, Jillian F

2017-03-07

Remediation of industrial wastewater is important for preventing environmental contamination and enabling water reuse. Biological treatment for one industrial contaminant, thiocyanate (SCN - ), relies upon microbial hydrolysis, but this process is sensitive to high loadings. To examine the activity and stability of a microbial community over increasing SCN - loadings, we established and operated a continuous-flow bioreactor fed increasing loadings of SCN - . A second reactor was fed ammonium sulfate to mimic breakdown products of SCN - . Biomass was sampled from both reactors for metagenomics and metaproteomics, yielding a set of genomes for 144 bacteria and one rotifer that constituted the abundant community in both reactors. We analyzed the metabolic potential and temporal dynamics of these organisms across the increasing loadings. In the SCN - reactor, Thiobacillus strains capable of SCN - degradation were highly abundant, whereas the ammonium sulfate reactor contained nitrifiers and heterotrophs capable of nitrate reduction. Key organisms in the SCN - reactor expressed proteins involved in SCN - degradation, sulfur oxidation, carbon fixation, and nitrogen removal. Lower performance at higher loadings was linked to changes in microbial community composition. This work provides an example of how meta-omics can increase our understanding of industrial wastewater treatment and inform iterative process design and development.

Application of microdosimetry on biological physics for ionizing radiation

NASA Astrophysics Data System (ADS)

Chen, Dandan; Sun, Liang

2018-02-01

Not Available Project supported by the National Natural Science Foundation of China (Grant Nos. 11304212 and 11575124), the Natural Science Foundation of Jiangsu Province, China (Grant Nos. BK20130279), the Priority Academic Program Development of Jiangsu Higher Education Institutions (PAPD), and the International Thermonuclear Experimental Reactor (ITER) Special Program of China (Grant No. 2014GB112006).

LOW-ENGINE-FRICTION TECHNOLOGY FOR ADVANCED NATURAL-GAS RECIPROCATING ENGINES

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

Victor W. Wong; Tian Tian; Grant Smedley

2003-08-28

This program aims at improving the efficiency of advanced natural-gas reciprocating engines (ANGRE) by reducing piston/ring assembly friction without major adverse effects on engine performance, such as increased oil consumption and emissions. A detailed set of piston/ring dynamic and friction models have been developed and applied that illustrated the fundamental relationships between design parameters and friction losses. Various low-friction strategies and concepts have been explored, and engine experiments will validate these concepts. An iterative process of experimentation, simulation and analysis, will be followed with the goal of demonstrating a complete optimized low-friction engine system. As planned, MIT has developed guidelinesmore » for an initial set of low-friction piston-ring-pack designs. Current recommendations focus on subtle top-piston-ring and oil-control-ring characteristics. A full-scale Waukesha F18 engine has been installed at Colorado State University and testing of the baseline configuration is in progress. Components for the first design iteration are being procured. Subsequent work includes examining the friction and engine performance data and extending the analyses to other areas to evaluate opportunities for further friction improvement and the impact on oil consumption/emission and wear, towards demonstrating an optimized reduced-friction engine system.« less

NIH-IEEE 2015 Strategic Conference on Healthcare Innovations and Point-of-Care Technologies for Prec

Cancer.gov

NIH and the Institute for Electrical and Electronics Engineering, Engineering in Medicine and Biology Society (IEEE/EMBS) hosted the third iteration of the Healthcare Innovations and Point-of-Care Technologies Conference last week.

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

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

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

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

ITER structural design criteria and their extension to advanced reactor blankets*1

NASA Astrophysics Data System (ADS)

Majumdar, S.; Kalinin, G.

2000-12-01

Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method.

ITER CS Model Coil and CS Insert Test Results

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

Martovetsky, N; Michael, P; Minervina, J

2000-09-07

The Inner and Outer modules of the Central Solenoid Model Coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting Central Solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t, OD is about 3.6 m and the stored energy is 640 MJ at 46 kA and peak field of 13 T. Testing of the CSMC and the CS Insert took place at Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the mainmore » results of the tests performed.« less

Simulation Modeling to Compare High-Throughput, Low-Iteration Optimization Strategies for Metabolic Engineering

PubMed Central

Heinsch, Stephen C.; Das, Siba R.; Smanski, Michael J.

2018-01-01

Increasing the final titer of a multi-gene metabolic pathway can be viewed as a multivariate optimization problem. While numerous multivariate optimization algorithms exist, few are specifically designed to accommodate the constraints posed by genetic engineering workflows. We present a strategy for optimizing expression levels across an arbitrary number of genes that requires few design-build-test iterations. We compare the performance of several optimization algorithms on a series of simulated expression landscapes. We show that optimal experimental design parameters depend on the degree of landscape ruggedness. This work provides a theoretical framework for designing and executing numerical optimization on multi-gene systems. PMID:29535690

A Huygens immersed-finite-element particle-in-cell method for modeling plasma-surface interactions with moving interface

NASA Astrophysics Data System (ADS)

Cao, Huijun; Cao, Yong; Chu, Yuchuan; He, Xiaoming; Lin, Tao

2018-06-01

Surface evolution is an unavoidable issue in engineering plasma applications. In this article an iterative method for modeling plasma-surface interactions with moving interface is proposed and validated. In this method, the plasma dynamics is simulated by an immersed finite element particle-in-cell (IFE-PIC) method, and the surface evolution is modeled by the Huygens wavelet method which is coupled with the iteration of the IFE-PIC method. Numerical experiments, including prototypical engineering applications, such as the erosion of Hall thruster channel wall, are presented to demonstrate features of this Huygens IFE-PIC method for simulating the dynamic plasma-surface interactions.

Software Estimates Costs of Testing Rocket Engines

NASA Technical Reports Server (NTRS)

Smith, C. L.

2003-01-01

Simulation-Based Cost Model (SiCM), a discrete event simulation developed in Extend , simulates pertinent aspects of the testing of rocket propulsion test articles for the purpose of estimating the costs of such testing during time intervals specified by its users. A user enters input data for control of simulations; information on the nature of, and activity in, a given testing project; and information on resources. Simulation objects are created on the basis of this input. Costs of the engineering-design, construction, and testing phases of a given project are estimated from numbers and labor rates of engineers and technicians employed in each phase, the duration of each phase; costs of materials used in each phase; and, for the testing phase, the rate of maintenance of the testing facility. The three main outputs of SiCM are (1) a curve, updated at each iteration of the simulation, that shows overall expenditures vs. time during the interval specified by the user; (2) a histogram of the total costs from all iterations of the simulation; and (3) table displaying means and variances of cumulative costs for each phase from all iterations. Other outputs include spending curves for each phase.

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

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

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

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

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

Jo, J.C.; Shin, W.K.; Choi, C.Y.

Transient heat transfer problems with phase changes (Stefan problems) occur in many engineering situations, including potential core melting and solidification during pressurized-water-reactor severe accidents, ablation of thermal shields, melting and solidification of alloys, and many others. This article addresses the numerical analysis of nonlinear transient heat transfer with melting or solidification. An effective and simple procedure is presented for the simulation of the motion of the boundary and the transient temperature field during the phase change process. To accomplish this purpose, an iterative implicit solution algorithm has been developed by employing the dual-reciprocity boundary-element method. The dual-reciprocity boundary-element approach providedmore » in this article is much simpler than the usual boundary-element method in applying a reciprocity principle and an available technique for dealing with the domain integral of the boundary element formulation simultaneously. In this article, attention is focused on two-dimensional melting (ablation)/solidification problems for simplicity. The accuracy and effectiveness of the present analysis method have been illustrated through comparisons of the calculation results of some examples of one-phase ablation/solidification problems with their known semianalytical or numerical solutions where available.« less

Identify and Quantify the Mechanistic Sources of Sensor Performance Variation Between Individual Sensors SN1 and SN2

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

Diaz, Aaron A.; Baldwin, David L.; Cinson, Anthony D.

2014-08-06

This Technical Letter Report satisfies the M3AR-14PN2301022 milestone, and is focused on identifying and quantifying the mechanistic sources of sensor performance variation between individual 22-element, linear phased-array sensor prototypes, SN1 and SN2. This effort constitutes an iterative evolution that supports the longer term goal of producing and demonstrating a pre-manufacturing prototype ultrasonic probe that possesses the fundamental performance characteristics necessary to enable the development of a high-temperature sodium-cooled fast reactor inspection system. The scope of the work for this portion of the PNNL effort conducted in FY14 includes performing a comparative evaluation and assessment of the performance characteristics of themore » SN1 and SN2 22 element PA-UT probes manufactured at PNNL. Key transducer performance parameters, such as sound field dimensions, resolution capabilities, frequency response, and bandwidth are used as a metric for the comparative evaluation and assessment of the SN1 and SN2 engineering test units.« less

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-03-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Metal Hall sensors for the new generation fusion reactors of DEMO scale

NASA Astrophysics Data System (ADS)

Bolshakova, I.; Bulavin, M.; Kargin, N.; Kost, Ya.; Kuech, T.; Kulikov, S.; Radishevskiy, M.; Shurygin, F.; Strikhanov, M.; Vasil'evskii, I.; Vasyliev, A.

2017-11-01

For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50-70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m-2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m-2 · s-1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (<1%). The results obtained make it possible to recommend gold sensors for magnetic diagnostics in the new generation fusion reactors of DEMO scale.

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

Vins, M.

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

Materials for DEMO and reactor applications—boundary conditions and new concepts

NASA Astrophysics Data System (ADS)

Coenen, J. W.; Antusch, S.; Aumann, M.; Biel, W.; Du, J.; Engels, J.; Heuer, S.; Houben, A.; Hoeschen, T.; Jasper, B.; Koch, F.; Linke, J.; Litnovsky, A.; Mao, Y.; Neu, R.; Pintsuk, G.; Riesch, J.; Rasinski, M.; Reiser, J.; Rieth, M.; Terra, A.; Unterberg, B.; Weber, Th; Wegener, T.; You, J.-H.; Linsmeier, Ch

2016-02-01

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here {{{W}}}{{f}}/{{W}} composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

150. ARAIII Reactor building (ARA608) Sections. Show highbay section, heater ...

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

150. ARA-III Reactor building (ARA-608) Sections. Show high-bay section, heater stack, and depth of reactor, piping, and heater pits. Aerojet-general 880-area/GCRE-608-A-3. Date: February 1958. Ineel index code no. 063-0608-00-013-102613. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

79. ARAIII. Early construction view of GCRE reactor building (ARA608) ...

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

79. ARA-III. Early construction view of GCRE reactor building (ARA-608) showing deep excavation, reinforcing steel, and forms for concrete placement for reactor and other pits. Camera facing southeast. July 22, 1958. Ineel photo no. 58-3466. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

ITER Cryoplant Infrastructures

NASA Astrophysics Data System (ADS)

Fauve, E.; Monneret, E.; Voigt, T.; Vincent, G.; Forgeas, A.; Simon, M.

2017-02-01

The ITER Tokamak requires an average 75 kW of refrigeration power at 4.5 K and 600 kW of refrigeration Power at 80 K to maintain the nominal operation condition of the ITER thermal shields, superconducting magnets and cryopumps. This is produced by the ITER Cryoplant, a complex cluster of refrigeration systems including in particular three identical Liquid Helium Plants and two identical Liquid Nitrogen Plants. Beyond the equipment directly part of the Cryoplant, colossal infrastructures are required. These infrastructures account for a large part of the Cryoplants lay-out, budget and engineering efforts. It is ITER Organization responsibility to ensure that all infrastructures are adequately sized and designed to interface with the Cryoplant. This proceeding presents the overall architecture of the cryoplant. It provides order of magnitude related to the cryoplant building and utilities: electricity, cooling water, heating, ventilation and air conditioning (HVAC).

Current status of nuclear engineering education

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

Palladino, N.J.

1975-09-01

The 65 colleges and universities offering undergraduate degrees in nuclear engineering and the 15 schools offering strong nuclear engineering options are, in general, doing a good job to meet the current spectrum of job opportunities. But, nuclear engineering programs are not producing enough graduates to meet growing demands. They currently receive little aid and support from their customers --industry and government--in the form of scholarships, grants, faculty research support, student thesis and project support, or student summer jobs. There is not enough interaction between industry and universities. Most nuclear engineering programs are geared too closely to the technology of themore » present family of reactors and too little to the future breeder reactors and controlled thermonuclear reactors. In addition, nuclear engineering programs attract too few women and members of minority ethnic groups. Further study of the reasons for this fact is needed so that effective corrective action can be taken. Faculty in nuclear engineering programs should assume greater initiative to provide attractive and objective nuclear energy electives for technical and nontechnical students in other disciplines to improve their technical understanding of the safety and environmental issues involved. More aggressive and persistent efforts must be made by nuclear engineering schools to obtain industry support and involvement in their programs. (auth)« less

Development of Modeling Approaches for Nuclear Thermal Propulsion Test Facilities

NASA Technical Reports Server (NTRS)

Jones, Daniel R.; Allgood, Daniel C.; Nguyen, Ke

2014-01-01

High efficiency of rocket propul-sion systems is essential for humanity to venture be-yond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rock-ets with relatively high thrust and twice the efficiency of the Space Shuttle Main Engine. NASA is in the pro-cess of developing a new NTP engine, and is evaluat-ing ground test facility concepts that allow for the thor-ough testing of NTP devices. NTP engine exhaust, hot gaseous hydrogen, is nominally expected to be free of radioactive byproducts from the nuclear reactor; how-ever, it has the potential to be contaminated due to off-nominal engine reactor performance. Several options are being investigated to mitigate this hazard potential with one option in particular that completely contains the engine exhaust during engine test operations. The exhaust products are subsequently disposed of between engine tests. For this concept (see Figure 1), oxygen is injected into the high-temperature hydrogen exhaust that reacts to produce steam, excess oxygen and any trace amounts of radioactive noble gases released by off-nominal NTP engine reactor performance. Water is injected to condense the potentially contaminated steam into water. This water and the gaseous oxygen (GO2) are subsequently passed to a containment area where the water and GO2 are separated into separate containment tanks.

From Intent to Action: An Iterative Engineering Process

ERIC Educational Resources Information Center

Mouton, Patrice; Rodet, Jacques; Vacaresse, Sylvain

2015-01-01

Quite by chance, and over the course of a few haphazard meetings, a Master's degree in "E-learning Design" gradually developed in a Faculty of Economics. Its original and evolving design was the result of an iterative process carried out, not by a single Instructional Designer (ID), but by a full ID team. Over the last 10 years it has…

PBF Reactor Building (PER620). PBF crane holds fuel test assembly ...

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

PBF Reactor Building (PER-620). PBF crane holds fuel test assembly aloft prior to lowering into reactor for test. Date: 1982. INEEL negative no. 82-4909 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

80. ARAIII. Forming of the mechanical equipment pit in reactor ...

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

80. ARA-III. Forming of the mechanical equipment pit in reactor building (ARA-608). Camera facing northwest. September 22, 1958. Ineel photo no. 58-4675. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

24. ARAIII Reactor building ARA608 interior. Camera facing south. Chalk ...

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

24. ARA-III Reactor building ARA-608 interior. Camera facing south. Chalk marks on wall indicate presence or absence of spot contamination. Ineel photo no. 3-2. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

Youinou, Gilles Jean-Michel

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment ofmore » advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E i), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀ ∞σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.« less

Task toward a Realization of Commercial Tokamak Fusion Plants in 2050 -The Role of ITER and the Succeeding Developments- 4.Technology and Material Research in Fusion Power Plant Development

NASA Astrophysics Data System (ADS)

Akiba, Masato; Matsui, Hideki; Takatsu, Hideyuki; Konishi, Satoshi

Technical issues regarding the fusion power plant that are required to be developed in the period of ITER construction and operation, both with ITER and with other facilities that complement ITER are described in this section. Three major fields are considered to be important in fusion technology. Section 4.1 summarizes blanket study, and ITER Test Blanket Module (TBM) development that focuses its effort on the first generation power blanket to be installed in DEMO. ITER will be equipped with 6 TBMs which are developed under each party's fusion program. In Japan, the solid breeder using water as a coolant is the primary candidate, and He-cooled pebble bed is the alternative. Other liquid options such as LiPb, Li or molten salt are developed by other parties' initiatives. The Test Blanket Working Group (TBWG) is coordinating these efforts. Japanese universities are investigating advanced concepts and fundamental crosscutting technologies. Section 4.2 introduces material development and particularly, the international irradiation facility, IFMIF. Reduced activation ferritic/martensitic steels are identified as promising candidates for the structural material of the first generation fusion blanket, while and vanadium alloy and SiC/SiC composite are pursued as advanced options. The IFMIF is currently planning the next phase of joint activity, EVEDA (Engineering Validation and Engineering Design Activity) that encompasses construction. Material studies together with the ITER TBM will provide essential technical information for development of the fusion power plant. Other technical issues to be addressed regarding the first generation fusion power plant are summarized in section 4.3. Development of components for ITER made remarkable progress for the major essential technology also necessary for future fusion plants, however many still need further improvements toward power plant. Such areas includes; the divertor, plasma heating/current drive, magnets, tritium, and remote handling. There remain many other technical issues for power plant which require integrated efforts.

Clean catalytic combustor program

NASA Technical Reports Server (NTRS)

Ekstedt, E. E.; Lyon, T. F.; Sabla, P. E.; Dodds, W. J.

1983-01-01

A combustor program was conducted to evolve and to identify the technology needed for, and to establish the credibility of, using combustors with catalytic reactors in modern high-pressure-ratio aircraft turbine engines. Two selected catalytic combustor concepts were designed, fabricated, and evaluated. The combustors were sized for use in the NASA/General Electric Energy Efficient Engine (E3). One of the combustor designs was a basic parallel-staged double-annular combustor. The second design was also a parallel-staged combustor but employed reverse flow cannular catalytic reactors. Subcomponent tests of fuel injection systems and of catalytic reactors for use in the combustion system were also conducted. Very low-level pollutant emissions and excellent combustor performance were achieved. However, it was obvious from these tests that extensive development of fuel/air preparation systems and considerable advancement in the steady-state operating temperature capability of catalytic reactor materials will be required prior to the consideration of catalytic combustion systems for use in high-pressure-ratio aircraft turbine engines.

PLUG STORAGE BUILDING, TRA611, AWAITS SHIELDING SOIL TO BE PLACED ...

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

PLUG STORAGE BUILDING, TRA-611, AWAITS SHIELDING SOIL TO BE PLACED OVER PLUG STORAGE TUBES. WING WALLS WILL SUPPORT EARTH FILL. MTR, PROCESS WATER BUILDING, AND WORKING RESERVOIR IN VIEW BEYOND PLUG STORAGE. CAMERA FACES NORTHEAST. INL NEGATIVE NO. 2949. Unknown Photographer, 7/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Development and Application of 3D Printed Mesoreactors in Chemical Engineering Education

ERIC Educational Resources Information Center

Tabassum, Tahseen; Iloska, Marija; Scuereb, Daniel; Taira, Noriko; Jin, Chongguang; Zaitsev, Vladimir; Afshar, Fara; Kim, Taejin

2018-01-01

3D printing technology has an enormous potential to apply to chemical engineering education. In this paper, we describe several designs of 3D printed mesoreactors (Y-shape, T-shape, and Long channel shape) using the following steps: reactor sketching, CAD modeling, and reactor printing. With a focus on continuous plug flow mesoreactors (PFRs, i.d.…

ELECTRICAL LINES ARRIVE FROM CENTRAL FACILITIES AREA, SOUTH OF MTR. ...

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

ELECTRICAL LINES ARRIVE FROM CENTRAL FACILITIES AREA, SOUTH OF MTR. EXCAVATION RUBBLE IN FOREGROUND. CONTRACTOR CRAFT SHOPS, CRANES, AND OTHER MATERIALS ON SITE. CAMERA FACES EAST, WITH LITTLE BUTTE AND MIDDLE BUTTE IN DISTANCE. INL NEGATIVE NO. 335. Unknown Photographer, 7/1/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR WING, TRA604. FIRST FLOOR PLAN. ENTRY LOBBY, MACHINE SHOP, ...

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

MTR WING, TRA-604. FIRST FLOOR PLAN. ENTRY LOBBY, MACHINE SHOP, INSTRUMENT SHOP, COUNTING ROOM, HEALTH PHYSICS LAB, LABS AND OFFICES, STORAGE, SHIPPING AND RECEIVING. BLAW-KNOX 3150-4-2, 7/1950. INL INDEX NO. 053-604-00-099-100008, REV. 7. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. EAST SIDE. CAMERA FACING WEST. ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. EAST SIDE. CAMERA FACING WEST. NOTE COURSE OF PIPE FROM GROUND AND FOLLOWING ROOF OF BUILDING. MTR BUILDING IN BACKGROUND AT RIGHT EDGE OF VIEW. INL NEGATIVE NO. HD46-36-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PRECONSTRUCTION IMAGE OF THE MTR SITE. ABANDONED IRRIGATION CANAL (FROM ...

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

PRE-CONSTRUCTION IMAGE OF THE MTR SITE. ABANDONED IRRIGATION CANAL (FROM EARLY 1900s) ILLUSTRATES FLATNESS OF MTR/TRA TERRAIN. FEATURE ON HORIZON IN LEFT OF VIEW IS EXPLORATORY WATER DRILLING EQUIPMENT. CAMERA LOOKS SOUTHEAST. INL NEGATIVE NO. 136. Unknown Photographer, 12/5/1949 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HOT CELL BUILDING, TRA632. FIRST FLOOR FOUNDATION PLAN SHOWS SECTIONALIZED ...

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

HOT CELL BUILDING, TRA-632. FIRST FLOOR FOUNDATION PLAN SHOWS SECTIONALIZED FLOOR LOADINGS AND CONCRETE SLAB THICKNESSES, A TYPICAL FEATURE OF NUCLEAR ARCHITECTURE. IDAHO OPERATIONS OFFICE MTR-632-IDO-2, 11/1952. INL INDEX NO. 531-0632-62-396-110561, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Comparison of reactivity in a flow reactor and a single cylinder engine

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

Natelson, Robert H.; Johnson, Rodney O.; Kurman, Matthew S.

2010-10-15

The relative reactivity of 2:1:1 and 1:1:1 mixtures of n-decane:n-butylcyclohexane:n-butylbenzene and an average sample of JP-8 were evaluated in a single cylinder engine and compared to results obtained in a pressurized flow reactor. At compression ratios of 14:1, 15:1, and 16:1, inlet temperature of 500 K, inlet pressure of 0.1 MPa, equivalence ratio of 0.23, and engine speed of 800 RPM, the autoignition delay times were, from shortest to longest, the 2:1:1, followed by the 1:1:1, and then the JP-8. This order corresponded with recent results in a pressurized flow reactor, where the preignition oxidation chemistry was monitored at temperaturesmore » of 600-800 K, 0.8 MPa pressure, and an equivalence ratio of 0.30, and where the preignition reactivity from highest to lowest was the 2:1:1, followed by the 1:1:1, and the JP-8. This shows that the relative reactivity at low temperatures in the flow reactor tracks the autoignition tendencies in the engine for these particular fuels. (author) the computed experimental error. (author)« less

A Methodology for Improving Active Learning Engineering Courses with a Large Number of Students and Teachers through Feedback Gathering and Iterative Refinement

ERIC Educational Resources Information Center

Estévez-Ayres, Iria; Alario-Hoyos, Carlos; Pérez-Sanagustín, Mar; Pardo, Abelardo; Crespo-García, Raquel M.; Leony, Derick; Parada G., Hugo A.; Delgado-Kloos, Carlos

2015-01-01

In the last decade, engineering education has evolved in many ways to meet society demands. Universities offer more flexible curricula and put a lot of effort on the acquisition of professional engineering skills by the students. In many universities, the courses in the first years of different engineering degrees share program and objectives,…

139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

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

139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Nuclear Reactor Safety--The APS Submits its Report

ERIC Educational Resources Information Center

Physics Today, 1975

1975-01-01

Presents the summary section of the American Physical Society (APS) report on the safety features of the light-water reactor, reviews the design, construction, and operation of a reactor and outlines the primary engineered safety features. Summarizes the major recommendations of the study group. (GS)

Automatic reactor model synthesis with genetic programming.

PubMed

Dürrenmatt, David J; Gujer, Willi

2012-01-01

Successful modeling of wastewater treatment plant (WWTP) processes requires an accurate description of the plant hydraulics. Common methods such as tracer experiments are difficult and costly and thus have limited applicability in practice; engineers are often forced to rely on their experience only. An implementation of grammar-based genetic programming with an encoding to represent hydraulic reactor models as program trees should fill this gap: The encoding enables the algorithm to construct arbitrary reactor models compatible with common software used for WWTP modeling by linking building blocks, such as continuous stirred-tank reactors. Discharge measurements and influent and effluent concentrations are the only required inputs. As shown in a synthetic example, the technique can be used to identify a set of reactor models that perform equally well. Instead of being guided by experience, the most suitable model can now be chosen by the engineer from the set. In a second example, temperature measurements at the influent and effluent of a primary clarifier are used to generate a reactor model. A virtual tracer experiment performed on the reactor model has good agreement with a tracer experiment performed on-site.

Reactor technology assessment and selection utilizing systems engineering approach

NASA Astrophysics Data System (ADS)

Zolkaffly, Muhammed Zulfakar; Han, Ki-In

2014-02-01

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

Biochemical Engineering Fundamentals

ERIC Educational Resources Information Center

Bailey, J. E.; Ollis, D. F.

1976-01-01

Discusses a biochemical engineering course that is offered as part of a chemical engineering curriculum and includes topics that influence the behavior of man-made or natural microbial or enzyme reactors. (MLH)

High internal inductance for steady-state operation in ITER and a reactor

DOE PAGES

Ferron, John R.; Holcomb, Christopher T.; Luce, Timothy C.; ...

2015-06-26

Increased confinement and ideal stability limits at relatively high values of the internal inductance (more » $${{\\ell}_{i}}$$ ) have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. Normalized plasma pressure in the range appropriate for a reactor has been achieved in high elongation and triangularity double-null divertor discharges with $${{\\beta}_{\\text{N}}}\\approx 5$$ at $${{\\ell}_{i}}\\approx 1.3$$ , near the ideal $n=1$ kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at $${{\\beta}_{\\text{N}}}>5.5$$ . Confinement is above the H-mode level with $${{H}_{98\\left(\\text{y},2\\right)}}\\approx 1.8$$ . At $${{q}_{95}}\\approx 7.5$$ , the current is overdriven, with bootstrap current fraction $${{f}_{\\text{BS}}}\\approx 0.8$$ , noninductive current fraction $${{f}_{\\text{NI}}}>1$$ and negative surface voltage. For ITER (which has a single-null divertor shape), operation at $${{\\ell}_{i}}\\approx 1$$ is a promising option with $${{f}_{\\text{BS}}}\\approx 0.5$$ and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at $${{q}_{95}}=4.8$$ , so far reaching $${{f}_{\\text{NI}}}=0.7$$ and $${{f}_{\\text{BS}}}=0.4$$ at $${{\\beta}_{\\text{N}}}\\approx 3.5$$ with performance appropriate for the ITER Q=5 mission, $${{H}_{89}}{{\\beta}_{\\text{N}}}/q_{95}^{2}\\approx 0.3$$ . Modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high $${{\\ell}_{i}}$$ discharge. Lastly, stable solutions in the double-null shape are found without the vacuum vessel wall at $${{\\beta}_{\\text{N}}}=4$$ , $${{\\ell}_{i}}=1.07$$ and $${{f}_{\\text{BS}}}=0.5$$ , and at $${{\\beta}_{\\text{N}}}=5$$ with the vacuum vessel wall.« less

DIII-D research to address key challenges for ITER and fusion energy

NASA Astrophysics Data System (ADS)

Buttery, R. J.; the DIII-D Team

2015-10-01

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modelling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelength turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully noninductively with βN = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a βN = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behaviour. Scenarios are shown to be compatible with radiative and snowflake divertor techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. Future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.

PROGRESS IN DESIGN OF THE INSTRUMENTATION AND CONTROL OF THE TOKAMAK COOLING WATER SYSTEM

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

Korsah, Kofi; DeVan, Bill; Ashburn, David

This paper discusses progress in the design of the control, interlock and safety systems of the Tokamak Cooling Water System (TCWS) for the ITER fusion reactor. The TCWS instrumentation and control (I&C) is one of approximately 200 separate plant I&C systems (e.g., vacuum system I&C, magnets system I&C) that interface to a common central I&C system through standardized networks. Several aspects of the I&C are similar to the I&C of fission-based power plants. However, some of the unique features of the ITER fusion reactor and the TCWS (e.g., high quasi-static magnetic field, need for baking and drying as well asmore » cooling operations), also demand some unique safety and qualification considerations. The paper compares the design strategy/guidelines of the TCWS I&C and the I&C of conventional nuclear power plants. Issues such as safety classifications, independence between control and safety systems, sensor sharing, redundancy, voting schemes, and qualification methodologies are discussed. It is concluded that independence and separation requirements are similar in both designs. However, the voting schemes for safety systems in nuclear power plants typically use 2oo4 (i.e., 4 divisions of safety I&C, any 2 of which is sufficient to trigger a safety action), while 2oo3 voting logic - within each of 2 independent trains - is used in the TCWS I&C. It is also noted that 2oo3 voting is also acceptable in nuclear power plants if adequate risk assessment and reliability is demonstrated. Finally, while qualification requirements provide similar guidance [e.g., both IEC 60780 (invoked in ITER-space), and IEEE 323 (invoked in fission power plant space) provide similar guidance], an important qualification consideration is the susceptibility of I&C to the magnetic fields of ITER. Also, the radiation environments are different. In the case of magnetic fields the paper discusses some options that are being considered.« less

DIII-D research to address key challenges for ITER and fusion energy

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

Buttery, Richard J.

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modeling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelengthmore » turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully non inductively with β N = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a β N = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behavior. Scenarios are shown to be compatible with radiative and snowflake diverter techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. As a result, future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.« less

DIII-D research to address key challenges for ITER and fusion energy

DOE PAGES

Buttery, Richard J.

2015-07-29

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modeling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelengthmore » turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully non inductively with β N = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a β N = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behavior. Scenarios are shown to be compatible with radiative and snowflake diverter techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. As a result, future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.« less

MTR,TRA603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED ...

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

MTR,TRA-603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED CUBICLES WERE IDENTIFIED BY SPONSORING LABORATORY AND ITS TEST HOLE NUMBER IN THE REACTOR, IE, "KAPL HB-1" SIGNIFIED KNOLLS ATOMIC POWER LABORATORY, HORIZONTAL BEAM NO. 1. "WAPD" WAS WESTINGHOUSE ATOMIC POWER DIVISION. CATCH TANKS AND SAMPLE STATIONS FOR TEST LOOPS WERE ASSOCIATED WITH THESE CUBICLES. NOTE DESKS, STORAGE CABINETS, SWITCH GEAR, INSTRUMENT PANELS. PHILLIPS PETROLEUM COMPANY MTR-E-5205, 4/1963. INL INDEX NO. 531-0603-00-706-009757, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

ETR, TRA642, CAMERA IS BELOW, BUT NEAR THE CEILING OF ...

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

ETR, TRA-642, CAMERA IS BELOW, BUT NEAR THE CEILING OF THE GROUND FLOOR, AND LOOKS DOWN TOWARD THE CONSOLE FLOOR. CAMERA FACES WESTERLY. THE REACTOR PIT IS IN THE CENTER OF THE VIEW. BEYOND IT TO THE LEFT IS THE SOUTH SIDE OF THE WORKING CANAL. IN THE FOREGROUND ON THE RIGHT IS THE SHIELDING FOR THE PROCESS WATER TUNNEL AND PIPING. SPIRAL STAIRCASE AT LEFT OF VIEW. INL NEGATIVE NO. 56-2237. Jack L. Anderson, Photographer, 7/6/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

148. ARAIII Reactor building (ARA608) Floor plan. Shows location of ...

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

148. ARA-III Reactor building (ARA-608) Floor plan. Shows location of reactor, heater, and mechanical loop pits; mechanical and electrical equipment rooms; and other work areas. Aerojet-general 880-area/GCRE-608-A-1. Date: February 1958. Ineel index code no. 063-0608-00-013-102612. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

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

158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

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

PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Assessment of the importance of neutron multiplication for tritium production

NASA Astrophysics Data System (ADS)

Chiovaro, P.; Di Maio, P. A.

2017-01-01

One of the major requirements for a fusion power plant in the future is tritium self-sufficiency. For this reason the scientific community has dedicated a lot of effort to research activity on reactor tritium breeding blankets. In the framework of the international project DEMO, many concepts of breeding blanket have been taken into account and some of them will be tested in the experimental reactor ITER by means of appropriate test blanket modules (TBMs). All the breeding blanket concepts rely on the adoption of binary systems composed of a material acting as neutronic multiplier and another as a breeder. This paper addresses a neutronic feature of these kinds of systems. In particular, attention has been focused on the assessment of the importance of neutrons coming from multiplication reactions for the production of tritium. A theoretical framework has been set up and a procedure to evaluate the performance of the multiplier-breeder systems, under the aforementioned point of view, has been developed. Moreover, the model set up has been applied to helium cooled lithium lead and helium cooled pebble bad TBMs under irradiation in ITER and the results have been critically discussed.

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and inmore » International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.« less

Experience Gained during the Adaptation of Classical ChE Subjects to the Bologna Plan in Europe: The Case of Chemical Reactors

ERIC Educational Resources Information Center

Ponsa, Sergio; Sanchez, Antoni

2011-01-01

At present, due to the overall adaptation to the European Higher Education Area (EHEA), a new concept regarding the teaching methodology was thought to be essential for engineering subjects. In this paper we describe our experience teaching the altered content of the courses on two classical subjects; Chemical Reactors (Chemical Engineering) and…

ETR, TRA642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST ...

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

ETR, TRA-642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST QUADRANT OF FLOOR. WESTINGHOUSE ATOMIC POWER DIVISION (WAPD) AND BETTIS ATOMIC POWER LABORATORY (BAPL) CONSUME MOST OF THE QUADRANT. PHILLIPS PETROLEUM COMPANY ETR-E-2256, 12/1966. INL INDEX NO. 532-0642-00-706-021256, REV. F. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. METAL FRAME OF BUILDING GOES ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. METAL FRAME OF BUILDING GOES UP IN BACKGROUND AS WORKERS PLACE A SECTION OF WATER LINE THAT WILL CARRY SECONDARY COOLANT BETWEEN HEAT EXCHANGER BUILDING AND THE COOLING TOWER. INL NEGATIVE NO. 56-2205. Jack L. Anderson, Photographer, 6/28/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. A PRIMARY COOLANT PUMP AND ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. A PRIMARY COOLANT PUMP AND 24-INCH CHECK VALVE ARE MOUNTED IN A SHIELDED CUBICLE. NOTE CONNECTION AT RIGHT THROUGH SHIELD WALL TO PUMP MOTOR ON OTHER SIDE. INL NEGATIVE NO. 56-4177. Jack L. Anderson, Photographer, 12/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR COMPRESSOR BUILDING, TRA643. CAMERA FACES NORTHEAST. WATER HEAT EXCHANGER ...

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

ETR COMPRESSOR BUILDING, TRA-643. CAMERA FACES NORTHEAST. WATER HEAT EXCHANGER IS IN LEFT FOREGROUND. A PARTIALLY ASSEMBLED PLANT AIR CONDITIONER IS AT CENTER. WORKERS AT RIGHT ASSEMBLE 4000 HORSEPOWER COMPRESSOR DRIVE MOTOR AT RIGHT. INL NEGATIVE NO. 56-3714. R.G. Larsen, Photographer, 11/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BUILDING INTERIOR, TRA603. BASEMENT. CAMERA IN WEST CORRIDOR FACING ...

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

MTR BUILDING INTERIOR, TRA-603. BASEMENT. CAMERA IN WEST CORRIDOR FACING SOUTH. FREIGHT ELEVATOR IS AT RIGHT OF VIEW. AT CENTER VIEW IS MTR VAULT NO. 1, USED TO STORE SPECIAL OR FISSIONABLE MATERIALS. INL NEGATIVE NO. HD46-6-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HOT CELL BUILDING, TRA632, INTERIOR. WRIGHT 3TON HOIST ON EAST ...

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

HOT CELL BUILDING, TRA-632, INTERIOR. WRIGHT 3-TON HOIST ON EAST SIDE OF CELL 2. SIGN AT LEFT OF VIEW SAYS, "...DO NOT BRING FISSILE MATERIAL INTO AREA WITHOUT APPROVAL." CAMERA FACES NORTHWEST. INL NEGATIVE NO. HD46-29-2. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR, TRA603. SECOND FLOOR PLAN. OFFICES AND INSTRUMENT ROOM. STEEL ...

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

MTR, TRA-603. SECOND FLOOR PLAN. OFFICES AND INSTRUMENT ROOM. STEEL PARTITIONS ON EAST SIDE OF INSTRUMENT ROOM. DETAIL OF COLUMN ENCASEMENTS. STAIRWAYS IN NORTH AND SOUTH CORNERS. PASSENGER ELEVATION. BLAW-KNOX 3150-803-3, 7/1950. INL INDEX NO. 531-0603-00-098-100562, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ADVANCED HEAT TRANSFER TEST FACILITY, TRA666A. ELEVATIONS. ROOF FRAMING PLAN. ...

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

ADVANCED HEAT TRANSFER TEST FACILITY, TRA-666A. ELEVATIONS. ROOF FRAMING PLAN. CONCRETE BLOCK SIDING. SLOPED ROOF. ROLL-UP DOOR. AIR INTAKE ENCLOSURE ON NORTH SIDE. F.C. TORKELSON 842-MTR-666-A5, 8/1966. INL INDEX NO. 531-0666-00-851-152258, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BUILDING, TRA603. DETAILED VIEW OF NORTHWEST CORNERS OF MTR ...

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

MTR BUILDING, TRA-603. DETAILED VIEW OF NORTHWEST CORNERS OF MTR HIGH-BAY AND SECOND/THIRD STORY SECTIONS. NOTE SHAPE OF PANEL ABOVE WINDOW OVER "TRA-603" BUILDING NUMBERS. THIS IS A "STANDARD PANEL." INL NEGATIVE NUMBER HD46-42-2. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLASH EVAPORATORS ARE PLACED ON UPPER ...

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

PROCESS WATER BUILDING, TRA-605. FLASH EVAPORATORS ARE PLACED ON UPPER LEVEL OF EAST SIDE OF BUILDING. WALLS WILL BE FORMED AROUND THEM. WORKING RESERVOIR BEYOND. CAMERA FACING EASTERLY. EXHAUST AIR STACK IS UNDER CONSTRUCTION AT RIGHT OF VIEW. INL NEGATIVE NO. 2579. Unknown Photographer, 6/18/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BASEMENT. WORKERS (DON ALVORD AND CYRIL VAN ORDEN OF ...

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

MTR BASEMENT. WORKERS (DON ALVORD AND CYRIL VAN ORDEN OF PHILLIPS PETROLEUM CO.) POSE FOR GAMMA IRRADIATION EXPERIMENT IN MTR CANAL. CANS OF FOOD WILL BE LOWERED TO CANAL BOTTOM, WHERE SPENT MTR FUEL ELEMENTS EMIT GAMMA RADIATION. INL NEGATIVE NO. 11746. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

CORNER OF SUBPILE ROOM: NORTH AND EAST SIDES. STEEL OUTER ...

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

CORNER OF SUBPILE ROOM: NORTH AND EAST SIDES. STEEL OUTER SHELL HAS BEEN AFFIXED. SIGN SAYS "HERRICK IRON WORKS STEEL, OAKLAND, CALIFORNIA." NOTE CONDUIT FOR FUTURE INSTRUMENTATION. TOP OF STEEL CASE WILL BE LEVEL WITH BASEMENT CEILING. CAMERA FACES SOUTHEAST. INL NEGATIVE NO. 734. Unknown Photographer, 10/6/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

ETR WASTE GAS EXITED THE ETR COMPLEX FROM THE NORTH ...

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

ETR WASTE GAS EXITED THE ETR COMPLEX FROM THE NORTH SIDE THROUGH A TUNNEL AND THEN TO A FILTER PIT. TUNNEL EXIT IS UNDER CONSTRUCTION WHILE CONTROL BUILDING IS BEING FORMED BEYOND. CAMERA FACING WEST. INL NEGATIVE NO. 56-1238. Jack L. Anderson, Photographer, 4/17/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Nonperturbative measurement of the local magnetic field using pulsed polarimetry for fusion reactor conditions (invited).

PubMed

Smith, Roger J

2008-10-01

A novel diagnostic technique for the remote and nonperturbative sensing of the local magnetic field in reactor relevant plasmas is presented. Pulsed polarimetry [Patent No. 12/150,169 (pending)] combines optical scattering with the Faraday effect. The polarimetric light detection and ranging (LIDAR)-like diagnostic has the potential to be a local B(pol) diagnostic on ITER and can achieve spatial resolutions of millimeters on high energy density (HED) plasmas using existing lasers. The pulsed polarimetry method is based on nonlocal measurements and subtle effects are introduced that are not present in either cw polarimetry or Thomson scattering LIDAR. Important features include the capability of simultaneously measuring local T(e), n(e), and B(parallel) along the line of sight, a resiliency to refractive effects, a short measurement duration providing near instantaneous data in time, and location for real-time feedback and control of magnetohydrodynamic (MHD) instabilities and the realization of a widely applicable internal magnetic field diagnostic for the magnetic fusion energy program. The technique improves for higher n(e)B(parallel) product and higher n(e) and is well suited for diagnosing the transient plasmas in the HED program. Larger devices such as ITER and DEMO are also better suited to the technique, allowing longer pulse lengths and thereby relaxing key technology constraints making pulsed polarimetry a valuable asset for next step devices. The pulsed polarimetry technique is clarified by way of illustration on the ITER tokamak and plasmas within the magnetized target fusion program within present technological means.

Establishing Physical and Engineering Science Base to Bridge from ITER to Demo

NASA Astrophysics Data System (ADS)

Peng, Y.-K. Martin; Abdou, M.; Gates, D.; Hegna, C.; Hill, D.; Najmabadi, F.; Navratil, G.; Parker, R.

2007-11-01

A Nuclear Component Testing (NCT) Discussion Group emerged recently to clarify how ``a lowered-risk, reduced-cost approach can provide a progressive fusion environment beyond the ITER level to explore, discover, and help establish the remaining, critically needed physical and engineering sciences knowledge base for Demo.'' The group, assuming success of ITER and other contemporary projects, identified critical ``gap-filling'' investigations: plasma startup, tritium self-sufficiency, plasma facing surface performance and maintainability, first wall/blanket/divertor materials defect control and lifetime management, and remote handling. Only standard or spherical tokamak plasma conditions below the advanced regime are assumed to lower the anticipated physics risk to continuous operation (˜2 weeks). Modular designs and remote handling capabilities are included to mitigate the risk of component failure and ease replacement. Aspect ratio should be varied to lower the cost, accounting for the contending physics risks and the near-term R&D. Cost and time-effective staging from H-H, D-D, to D-T will also be considered. *Work supported by USDOE.

Evaluating the iterative development of VR/AR human factors tools for manual work.

PubMed

Liston, Paul M; Kay, Alison; Cromie, Sam; Leva, Chiara; D'Cruz, Mirabelle; Patel, Harshada; Langley, Alyson; Sharples, Sarah; Aromaa, Susanna

2012-01-01

This paper outlines the approach taken to iteratively evaluate a set of VR/AR (virtual reality / augmented reality) applications for five different manual-work applications - terrestrial spacecraft assembly, assembly-line design, remote maintenance of trains, maintenance of nuclear reactors, and large-machine assembly process design - and examines the evaluation data for evidence of the effectiveness of the evaluation framework as well as the benefits to the development process of feedback from iterative evaluation. ManuVAR is an EU-funded research project that is working to develop an innovative technology platform and a framework to support high-value, high-knowledge manual work throughout the product lifecycle. The results of this study demonstrate the iterative improvements reached throughout the design cycles, observable through the trending of the quantitative results from three successive trials of the applications and the investigation of the qualitative interview findings. The paper discusses the limitations of evaluation in complex, multi-disciplinary development projects and finds evidence of the effectiveness of the use of the particular set of complementary evaluation methods incorporating a common inquiry structure used for the evaluation - particularly in facilitating triangulation of the data.

Evaluation of power transfer efficiency for a high power inductively coupled radio-frequency hydrogen ion source

NASA Astrophysics Data System (ADS)

Jain, P.; Recchia, M.; Cavenago, M.; Fantz, U.; Gaio, E.; Kraus, W.; Maistrello, A.; Veltri, P.

2018-04-01

Neutral beam injection (NBI) for plasma heating and current drive is necessary for International Thermonuclear Experimental reactor (ITER) tokamak. Due to its various advantages, a radio frequency (RF) driven plasma source type was selected as a reference ion source for the ITER heating NBI. The ITER relevant RF negative ion sources are inductively coupled (IC) devices whose operational working frequency has been chosen to be 1 MHz and are characterized by high RF power density (˜9.4 W cm-3) and low operational pressure (around 0.3 Pa). The RF field is produced by a coil in a cylindrical chamber leading to a plasma generation followed by its expansion inside the chamber. This paper recalls different concepts based on which a methodology is developed to evaluate the efficiency of the RF power transfer to hydrogen plasma. This efficiency is then analyzed as a function of the working frequency and in dependence of other operating source and plasma parameters. The study is applied to a high power IC RF hydrogen ion source which is similar to one simplified driver of the ELISE source (half the size of the ITER NBI source).

Reliability and safety of the electrical power supply complex of the Hanford production reactors

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

Robbins, F.D.

Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high ordermore » of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.« less

Chemical Reactors.

ERIC Educational Resources Information Center

Kenney, C. N.

1980-01-01

Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

IDC Re-Engineering Phase 2 Glossary Version 1.3

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

Young, Christopher J.; Harris, James M.

2017-01-01

This document contains the glossary of terms used for the IDC Re-Engineering Phase 2 project. This version was created for Iteration E3. The IDC applies automatic processing methods in order to produce, archive, and distribute standard IDC products on behalf of all States Parties.

In-vessel tritium retention and removal in ITER

NASA Astrophysics Data System (ADS)

Federici, G.; Anderl, R. A.; Andrew, P.; Brooks, J. N.; Causey, R. A.; Coad, J. P.; Cowgill, D.; Doerner, R. P.; Haasz, A. A.; Janeschitz, G.; Jacob, W.; Longhurst, G. R.; Nygren, R.; Peacock, A.; Pick, M. A.; Philipps, V.; Roth, J.; Skinner, C. H.; Wampler, W. R.

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER - and more generically in any other next-step experimental fusion facility fuelled with tritium - the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

NASA Astrophysics Data System (ADS)

Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

2009-12-01

Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

Experience on divertor fuel retention after two ITER-Like Wall campaigns

NASA Astrophysics Data System (ADS)

Heinola, K.; Widdowson, A.; Likonen, J.; Ahlgren, T.; Alves, E.; Ayres, C. F.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Guillemaut, C.; Jepu, I.; Krat, S.; Lahtinen, A.; Matthews, G. F.; Mayer, M.; Contributors, JET

2017-12-01

The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during the operation of a future fusion reactor. Present work summarizes and reports the plasma fuel retention in the divertor resulting from the two first experimental campaigns with the ITER-Like Wall. The deposition pattern in the divertor after the second campaign shows same trend as was observed after the first campaign: highest deposition of 10-15 μm was found on the top part of the inner divertor. Due to the change in plasma magnetic configurations from the first to the second campaign, and the resulted strike point locations, an increase of deposition was observed on the base of the divertor. The deuterium retention was found to be affected by the hydrogen plasma experiments done at the end of second experimental campaign.

87. ARAIII. GCRE reactor building (ARA608) Mechanical equipment room. Utility ...

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

87. ARA-III. GCRE reactor building (ARA-608) Mechanical equipment room. Utility air receiver, dryer, and compressor sit on their foundations prior to grouting. December 22, 1958. Ineel photo no. 58-6429. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

65. ARAII. Interior view of SL1 reactor building control piping ...

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

65. ARA-II. Interior view of SL-1 reactor building control piping for water purification system. On operating floor of building. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

154. ARAIII Reactor building (ARA608) Foundation sections and details. Shows ...

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

154. ARA-III Reactor building (ARA-608) Foundation sections and details. Shows profiles of pits. Aerojet-general 888-area/GCRE-608-S-2. Date: February 1958. Ineel index code no. 062-0608-60-013-102654. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

153. ARAIII Reactor building (ARA608) Foundation plan. Aerojetgeneral 880area/GCRE608S1. Date: ...

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

153. ARA-III Reactor building (ARA-608) Foundation plan. Aerojet-general 880-area/GCRE-608-S-1. Date: February 1958. Ineel index code no. 063-0608-60-013-102653. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

152. ARAIII Reactor building (ARA608) Details of heater and piping ...

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

152. ARA-III Reactor building (ARA-608) Details of heater and piping pits, including instrumentation plan. Aerojet-general 880-area/GCRE-608-T-18. Date: November 1958. Ineel index code no. 063-0608-25-013-102677. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

149. ARAIII Reactor building (ARA608) Exterior elevations, showing north, south, ...

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

149. ARA-III Reactor building (ARA-608) Exterior elevations, showing north, south, east, and west. Aerojet-general 880-area/GCRE-608-A-6. Date: February 1958. Ineel index code no. 063-0608-00-013-102615. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Partial oxidation for improved cold starts in alcohol-fueled engines: Phase 2 topical report

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

NONE

1998-04-01

Alcohol fuels exhibit poor cold-start performance because of their low volatility. Neat alcohol engines become difficult, if not impossible, to start at temperatures close to or below freezing. Improvements in the cold-start performance (both time to start and emissions) are essential to capture the full benefits of alcohols as an alternative transportation fuel. The objective of this project was to develop a neat alcohol partial oxidation (POX) reforming technology to improve an alcohol engine`s ability to start at low temperatures (as low as {minus}30 C) and to reduce its cold-start emissions. The project emphasis was on fuel-grade ethanol (E95) butmore » the technology can be easily extended to other alcohol fuels. Ultimately a compact, on-vehicle, ethanol POX reactor was developed as a fuel system component to produce a hydrogen-rich, fuel-gas mixture for cold starts. The POX reactor is an easily controllable combustion device that allows flexibility during engine startup even in the most extreme conditions. It is a small device that is mounted directly onto the engine intake manifold. The gaseous fuel products (or reformate) from the POX reactor exit the chamber and enter the intake manifold, either replacing or supplementing the standard ethanol fuel consumed during an engine start. The combustion of the reformate during startup can reduce engine start time and tail-pipe emissions.« less

PBF Reactor Building (PER620). Camera faces south toward verticallift door, ...

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

PBF Reactor Building (PER-620). Camera faces south toward vertical-lift door, which is closed. Note crane and its trolley positioned near door; its rails along side walls. Reactor vessel and lifting beams are positioned above reactor pit. Photographer: John Capek. Date: January 9, 1970. INEEL negative no. 70-132 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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

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

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

Simulations of carbon sputtering in fusion reactor divertor plates

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

Marian, J; Zepeda-Ruiz, L A; Gilmer, G H

2005-10-03

The interaction of edge plasma with material surfaces raises key issues for the viability of the International Thermonuclear Reactor (ITER) and future fusion reactors, including heat-flux limits, net material erosion, and impurity production. After exposure of the graphite divertor plate to the plasma in a fusion device, an amorphous C/H layer forms. This layer contains 20-30 atomic percent D/T bonded to C. Subsequent D/T impingement on this layer produces a variety of hydrocarbons that are sputtered back into the sheath region. We present molecular dynamics (MD) simulations of D/T impacts on amorphous carbon layer as a function of ion energymore » and orientation, using the AIREBO potential. In particular, energies are varied between 10 and 150 eV to transition from chemical to physical sputtering. These results are used to quantify yield, hydrocarbon composition and eventual plasma contamination.« less

Alloying of steel and graphite by hydrogen in nuclear reactor

NASA Astrophysics Data System (ADS)

Krasikov, E.

2017-02-01

In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

U. S. fusion programs: Struggling to stay in the game

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

Crawford, M.

Funding for the US fusion energy program has suffered and will probably continue to suffer major cuts. A committee hand-picked by Energy Secretary James Watkins urged the Department of Energy to mount an aggressive program to develop fusion power, but congress cut funding from $323 million in 1990 to $275 million in 1991. This portends dire conditions for fusion research and development. Projects to receive top priority are concerned with the tokamaks and to keep the next big machine, the Burning Plasma Experiment, scheduled for beginning of construction in 1993 on schedule. Secretary Watkins is said to want to keepmore » the International Thermonuclear Energy Reactor (ITER) on schedule. ITER would follow the Burning Plasma Experiment.« less

Genome-Resolved Meta-Omics Ties Microbial Dynamics to Process Performance in Biotechnology for Thiocyanate Degradation

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

Kantor, Rose S.; Huddy, Robert J.; Iyer, Ramsunder

Remediation of industrial wastewater is important for preventing environmental contamination and allowing water reuse. Biological treatment for one industrial contaminant, thiocyanate (SCN - ), relies upon microbial hydrolysis, but this process is sensitive to high loadings. To examine the activity and stability of a microbial community over increasing SCN - loadings, we established and operated a continuous-flow bioreactor fed increasing loadings of SCN - . A second reactor was fed ammonium sulfate to mimic breakdown products of SCN - . Biomass was sampled from both reactors for metagenomics and metaproteomics, yielding a set of genomes for 144 bacteria and onemore » rotifer that constituted the abundant community in both reactors. We analyzed the metabolic potential and temporal dynamics of these organisms across the increasing loadings. In the SCN - reactor, Thiobacillus strains capable of SCN - degradation were highly abundant, whereas the ammonium sulfate reactor contained nitrifiers and heterotrophs capable of nitrate reduction. Key organisms in the SCN - reactor expressed proteins involved in SCN - degradation, sulfur oxidation, carbon fixation, and nitrogen removal. Lower performance at higher loadings was linked to changes in microbial community composition. This work provides an example of how meta-omics can increase our understanding of industrial wastewater treatment and inform iterative process design and development.« less

Genome-Resolved Meta-Omics Ties Microbial Dynamics to Process Performance in Biotechnology for Thiocyanate Degradation

DOE PAGES

Kantor, Rose S.; Huddy, Robert J.; Iyer, Ramsunder; ...

2017-01-31

Remediation of industrial wastewater is important for preventing environmental contamination and allowing water reuse. Biological treatment for one industrial contaminant, thiocyanate (SCN - ), relies upon microbial hydrolysis, but this process is sensitive to high loadings. To examine the activity and stability of a microbial community over increasing SCN - loadings, we established and operated a continuous-flow bioreactor fed increasing loadings of SCN - . A second reactor was fed ammonium sulfate to mimic breakdown products of SCN - . Biomass was sampled from both reactors for metagenomics and metaproteomics, yielding a set of genomes for 144 bacteria and onemore » rotifer that constituted the abundant community in both reactors. We analyzed the metabolic potential and temporal dynamics of these organisms across the increasing loadings. In the SCN - reactor, Thiobacillus strains capable of SCN - degradation were highly abundant, whereas the ammonium sulfate reactor contained nitrifiers and heterotrophs capable of nitrate reduction. Key organisms in the SCN - reactor expressed proteins involved in SCN - degradation, sulfur oxidation, carbon fixation, and nitrogen removal. Lower performance at higher loadings was linked to changes in microbial community composition. This work provides an example of how meta-omics can increase our understanding of industrial wastewater treatment and inform iterative process design and development.« less

NUCLEAR REACTOR AS THE OBJECT OF CONTROL. AUTOMATIC CONTROL OF AIRCRAFT ENGINES . B.S. Voronkev Collection of Articles

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

None

BS> The dynamics of a power reactor is treated in some detail. Although the reactor is described by a nonlinear differential equation of the seventh order, a two-group approximstion with prompt neutrons and one averaged group of delayed neutrons may be used. When the reactor is in equilibrium, the reactor equation may be linearized in two ways. The effects of positive and negative coefficients of tins of the reactor are discussed. The nonlinear character of the control rods is trested. (D.L.C.)

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

The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

2013-10-01

A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

Teaching Engineering Design Through Paper Rockets

ERIC Educational Resources Information Center

Welling, Jonathan; Wright, Geoffrey A.

2018-01-01

The paper rocket activity described in this article effectively teaches the engineering design process (EDP) by engaging students in a problem-based learning activity that encourages iterative design. For example, the first rockets the students build typically only fly between 30 and 100 feet. As students test and evaluate their rocket designs,…

Leadership in Government Organization Change Efforts: A Multi-Case Analysis

DTIC Science & Technology

2004-09-01

cleared contractor employees on short notice, in the event of demand surges. Letting go of contractor employees was considered a safer bet than...principle can be seen in the early establishment of two reactor plant-engineering laboratories. The strong, competitive capabilities of these two...personnel. The list could extend indefinitely. Application of this principle can be seen in the early establishment of two reactor plant-engineering

TRITIUM LABORATORY, TRA666, INTERIOR. MAIN FLOOR. CONTROL ROOM ENCLOSURE AT ...

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

TRITIUM LABORATORY, TRA-666, INTERIOR. MAIN FLOOR. CONTROL ROOM ENCLOSURE AT CENTER OF VIEW. SIGN ABOVE DOOR SAYS "HYDRAULIC TEST FACILITY CONTROL ROOM." SIGN IN WINDOW SAYS "EATING AREA." "EVACUATION AND EMERGENCY INFORMATION" IS POSTED ON CABINET AT LEFT OF VIEW. INL NEGATIVE NO. HD30-2-3. Mike Crane, Photographer, 6/2001 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HEALTH AND SAFETY BUILDING, TRA667. SOUTH AND WEST ELEVATIONS. FLOOR ...

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

HEALTH AND SAFETY BUILDING, TRA-667. SOUTH AND WEST ELEVATIONS. FLOOR PLAN AND ROOM DESIGNATIONS. NOTE PAIR OF ENTRY DOORS IN WEST ELEVATION FOR MEN AND WOMEN. CONCRETE T-BEAMS. F.C. TORKELSON CO. 842-MTR-667-A1, 1/1963. INL INDEX NO. 531-0667-00-851-151143, REV. 4. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. CONTEXTUAL VIEW, CAMERA FACING SOUTHEAST. PROCESS ...

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

PROCESS WATER BUILDING, TRA-605. CONTEXTUAL VIEW, CAMERA FACING SOUTHEAST. PROCESS WATER BUILDING AND ETR STACK ARE IN LEFT HALF OF VIEW. TRA-666 IS NEAR CENTER, ABUTTED BY SECURITY BUILDING; TRA-626, AT RIGHT EDGE OF VIEW BEHIND BUS. INL NEGATIVE NO. HD46-34-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. ON GROUND FLOOR, CAMERA LOOKS SOUTHWEST INTO PIT. ...

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

ETR, TRA-642. ON GROUND FLOOR, CAMERA LOOKS SOUTHWEST INTO PIT. CANAL STRUCTURE IS AT RIGHT OF CENTER WITH RECTANGULAR OPENING TO BE MATED WITH THE DE-FUELING MECHANISM THAT WILL DEPOSIT FUEL RODS INTO THE WORKING CANAL. INL NEGATIVE NO. 56-3710. R.G. Larsen, Photographer, 11/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BUILDING, TRA603. EAST SIDE. CAMERA FACING WEST. CORRUGATED IRON ...

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

MTR BUILDING, TRA-603. EAST SIDE. CAMERA FACING WEST. CORRUGATED IRON BUILDING MARKED WITH "X" IS TRA-651. TRA-626, TO ITS RIGHT, HOUSED COMPRESSOR EQUIPMENT FOR THE AIRCRAFT NUCLEAR PROPULSION PROGRAM. LATER, IT WAS USED FOR STORAGE. INL NEGATIVE NO. HD46-42-4. Mike Crane, Photographer, April 2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BASEMENT, TRA642, INTERIOR. BASEMENT. CUBICLE INTERIOR (SEE PHOTOS ID33G101 ...

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

ETR BASEMENT, TRA-642, INTERIOR. BASEMENT. CUBICLE INTERIOR (SEE PHOTOS ID-33-G-101 AND ID-33-G-102) WITH TANK AND SODIUM-RELATED APPARATUS. CAMERA STANDS BEFORE ROLL-UP DOOR SHOWN IN PHOTO ID-33-G-101. INL NEGATIVE NO. HD24-3-3. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. SECTIONS B, C AND D SHOW ...

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

PROCESS WATER BUILDING, TRA-605. SECTIONS B, C AND D SHOW RELATIONSHIP BETWEEN FLASH EVAPORATORS (ABOVE) AND SEAL AND SUMP TANKS (BELOW). BASEMENT FLOOR IS BELOW GRADE; FIRST FLOOR, ABOVE GRADE. SHIELDING TOLERANCES. BLAW-KNOX 3150-5-7, 8/1950. INL INDEX NO. 531-605-00-098-100012, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. EASTWEST SECTION, LOOKING NORTH. PATH OF COOLING WATER ...

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

ETR, TRA-642. EAST-WEST SECTION, LOOKING NORTH. PATH OF COOLING WATER PIPE TUNNEL. WORKING AND STORAGE CANAL. SUB-PILE ROOM. CONTROL ROD ACCESS ROOM. FLOOR NAMES. (THIS WAS A CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-5, 11/1955. INL INDEX NO. 532-0642-00-486-100913. REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR BASEMENT. DOORWAY TO SOURCE STORAGE VAULT IS AT CENTER ...

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

MTR BASEMENT. DOORWAY TO SOURCE STORAGE VAULT IS AT CENTER OF VIEW; TO DECONTAMINATION ROOM, AT RIGHT. PART OF MAZE ENTRY IS VISIBLE INSIDE VAULT DOORWAY. INL NEGATIVE NO. 7763. Unknown Photographer, photo was dated as 3/30/1953, but this was probably an error. The more likely date is 3/30/1952. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HOT CELL BUILDING, TRA632, INTERIOR. WINDOWED ROOM IS OFFICE; NEXT ...

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

HOT CELL BUILDING, TRA-632, INTERIOR. WINDOWED ROOM IS OFFICE; NEXT DOOR WAS DARKROOM, AND THIRD DOOR LED TO ANOTHER OFFICE. ALL ARE ALONG NORTH WALL OF BUILDING (ETR EXTENSION OF 1958). CAMERA FACES NORTHEAST. PUMICE BLOCK WALLS. INL NEGATIVE NO. HD46-29-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Systems Engineering of Electric and Hybrid Vehicles

NASA Technical Reports Server (NTRS)

Kurtz, D. W.; Levin, R. R.

1986-01-01

Technical paper notes systems engineering principles applied to development of electric and hybrid vehicles such that system performance requirements support overall program goal of reduced petroleum consumption. Paper discusses iterative design approach dictated by systems analyses. In addition to obvious peformance parameters of range, acceleration rate, and energy consumption, systems engineering also considers such major factors as cost, safety, reliability, comfort, necessary supporting infrastructure, and availability of materials.

ETR, TRA642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY ...

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

ETR, TRA-642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY FROM TOP DECK OF COOLING TOWER. SHADOW IS CAST BY COOLING TOWER UNITS OFF LEFT OF VIEW. HIGH-BAY REACTOR BUILDING IS SURROUNDED BY ITS ATTACHED SERVICES: ELECTRICAL (TRA-648), HEAT EXCHANGER (TRA-644 WITH U-SHAPED YARD), AND COMPRESSOR (TRA-643). THE CONTROL BUILDING (TRA-647) ON THE NORTH SIDE IS HIDDEN FROM VIEW. AT UPPER RIGHT IS MTR BUILDING, TRA-603. INL NEGATIVE NO. 56-3798. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Engineered heat treated methanogenic granules: a promising biotechnological approach for extreme thermophilic biohydrogen production.

PubMed

Abreu, Angela A; Alves, Joana I; Pereira, M Alcina; Karakashev, Dimitar; Alves, M Madalena; Angelidaki, Irini

2010-12-01

In the present study, two granular systems were compared in terms of hydrogen production rate, stability and bacterial diversity under extreme thermophilic conditions (70 degrees C). Two EGSB reactors were individually inoculated with heat treated methanogenic granules (HTG) and HTG amended with enrichment culture with high capacity of hydrogen production (engineered heat treated methanogenic granules - EHTG), respectively. The reactor inoculated with EHTG (R(EHTG)) attained a maximum production rate of 2.7l H(2)l(-1)day(-1) in steady state. In comparison, the R(HTG) containing the HTG granules was very unstable, with low hydrogen productions and only two peaks of hydrogen (0.8 and 1.5l H(2)l(-1)day(-1)). The presence of active hydrogen producers in the R(EHTG) system during the reactor start-up resulted in the development of an efficient H(2)-producing bacterial community. The results showed that "engineered inocula" where known hydrogen producers are co-inoculated with HTG is an efficient way to start up biohydrogen-producing reactors. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

The hard start phenomena in hypergolic engines. Volume 4: The chemistry of hydrazine fuels and nitrogen tetroxide propellant systems

NASA Technical Reports Server (NTRS)

Miron, Y.; Perlee, H. E.

1974-01-01

The various chemical reactions that occur and that could possibly occur in the RCS engines utilizing hydrazine-type fuel/nitrogen tetroxide propellant systems, prior to ignition (preignition), during combustion, and after combustion (postcombustion), and endeavors to relate the hard-start phenomenon to some of these reactions are discussed. The discussion is based on studies utilizing a variety of experimental techniques and apparatus as well as current theories of chemical reactions and reaction kinetics. The chemical reactions were studied in low pressure gas flow reactors, low temperature homogeneous- and heterogeneous-phase reactors, simulated two-dimensional (2-D) engines, and scaled and full size engines.

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Programmore » since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.« less

The external kink mode in diverted tokamaks

NASA Astrophysics Data System (ADS)

Turnbull, A. D.; Hanson, J. M.; Turco, F.; Ferraro, N. M.; Lanctot, M. J.; Lao, L. L.; Strait, E. J.; Piovesan, P.; Martin, P.

2016-06-01

> . The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.

Fission-powered in-core thermoacoustic sensor

DOE PAGES

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.; ...

2016-04-07

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. Furthermore, these signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Fission-powered in-core thermoacoustic sensor

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

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.

2016-04-04

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

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

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

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

Development of a Twin-spool Turbofan Engine Simulation Using the Toolbox for Modeling and Analysis of Thermodynamic Systems (T-MATS)

NASA Technical Reports Server (NTRS)

Zinnecker, Alicia M.; Chapman, Jeffryes W.; Lavelle, Thomas M.; Litt, Johathan S.

2014-01-01

The Toolbox for Modeling and Analysis of Thermodynamic Systems (T-MATS) is a tool that has been developed to allow a user to build custom models of systems governed by thermodynamic principles using a template to model each basic process. Validation of this tool in an engine model application was performed through reconstruction of the Commercial Modular Aero-Propulsion System Simulation (C-MAPSS) (v2) using the building blocks from the T-MATS (v1) library. In order to match the two engine models, it was necessary to address differences in several assumptions made in the two modeling approaches. After these modifications were made, validation of the engine model continued by integrating both a steady-state and dynamic iterative solver with the engine plant and comparing results from steady-state and transient simulation of the T-MATS and C-MAPSS models. The results show that the T-MATS engine model was accurate within 3 of the C-MAPSS model, with inaccuracy attributed to the increased dimension of the iterative solver solution space required by the engine model constructed using the T-MATS library. This demonstrates that, given an understanding of the modeling assumptions made in T-MATS and a baseline model, the T-MATS tool provides a viable option for constructing a computational model of a twin-spool turbofan engine that may be used in simulation studies.

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

Batista, Antonio J. N.; Santos, Bruno; Fernandes, Ana

The data acquisition and control instrumentation cubicles room of the ITER tokamak will be irradiated with neutrons during the fusion reactor operation. A Virtex-6 FPGA from Xilinx (XC6VLX365T-1FFG1156C) is used on the ATCA-IO-PROCESSOR board, included in the ITER Catalog of I and C products - Fast Controllers. The Virtex-6 is a re-programmable logic device where the configuration is stored in Static RAM (SRAM), functional data stored in dedicated Block RAM (BRAM) and functional state logic in Flip-Flops. Single Event Upsets (SEU) due to the ionizing radiation of neutrons causes soft errors, unintended changes (bit-flips) to the values stored in statemore » elements of the FPGA. The SEU monitoring and soft errors repairing, when possible, were explored in this work. An FPGA built-in Soft Error Mitigation (SEM) controller detects and corrects soft errors in the FPGA configuration memory. Novel SEU sensors with Error Correction Code (ECC) detect and repair the BRAM memories. Proper management of SEU can increase reliability and availability of control instrumentation hardware for nuclear applications. The results of the tests performed using the SEM controller and the BRAM SEU sensors are presented for a Virtex-6 FPGA (XC6VLX240T-1FFG1156C) when irradiated with neutrons from the Portuguese Research Reactor (RPI), a 1 MW nuclear fission reactor operated by IST in the neighborhood of Lisbon. Results show that the proposed SEU mitigation technique is able to repair the majority of the detected SEU errors in the configuration and BRAM memories. (authors)« less

PBF Reactor Building (PER620). Camera is in cab of electricpowered ...

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

PBF Reactor Building (PER-620). Camera is in cab of electric-powered rail crane and facing east. Reactor pit and storage canal have been shaped. Floors for wings on east and west side are above and below reactor in view. Photographer: Larry Page. Date: August 23, 1967. INEEL negative no. 67-4403 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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.

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…

PBF Reactor Building (PER620). Camera in second basement near subpile ...

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

PBF Reactor Building (PER-620). Camera in second basement near sub-pile room (directly below reactor vessel). Door and penetrations lead to sub-pile room. Date: August 15, 1969. Photographer: Larry Page. INEEL negative no. 69-4310 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

90. ARAIII. GCRE reactor building (ARA608) mechanical loop pit. Shows ...

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

90. ARA-III. GCRE reactor building (ARA-608) mechanical loop pit. Shows nitrogen gas compressor in foreground, piping installations on walls of pit, and other details. February 24, 1959. Ineel photo no. 59-880. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

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

LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

LOFT. Reactor apparatus leaves A&M building (TAN607). Shielded locomotive has ...

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

LOFT. Reactor apparatus leaves A&M building (TAN-607). Shielded locomotive has aerojet logo, which replaced old general electric logo, pulls reactor from assembly shop on dolly. Camera facing easterly. Date: 1973. INEEL negative no. 73-3700 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Biocatalysis engineering: the big picture.

PubMed

Sheldon, Roger A; Pereira, Pedro C

2017-05-22

In this tutorial review we describe a holistic approach to the invention, development and optimisation of biotransformations utilising isolated enzymes. Increasing attention to applied biocatalysis is motivated by its numerous economic and environmental benefits. Biocatalysis engineering concerns the development of enzymatic systems as a whole, which entails engineering its different components: substrate engineering, medium engineering, protein (enzyme) engineering, biocatalyst (formulation) engineering, biocatalytic cascade engineering and reactor engineering.

Application of a Systems Engineering Approach to Support Space Reactor Development

NASA Astrophysics Data System (ADS)

Wold, Scott

2005-02-01

In 1992, approximately 25 Russian and 12 U.S. engineers and technicians were involved in the transport, assembly, inspection, and testing of over 90 tons of Russian equipment associated with the Thermionic System Evaluation Test (TSET) Facility. The entire Russian Baikal Test Stand, consisting of a 5.79 m tall vacuum chamber and related support equipment, was reassembled and tested at the TSET facility in less than four months. In November 1992, the first non-nuclear operational test of a complete thermionic power reactor system in the U.S. was accomplished three months ahead of schedule and under budget. A major factor in this accomplishment was the application of a disciplined top-down systems engineering approach and application of a spiral development model to achieve the desired objectives of the TOPAZ International Program (TIP). Systems Engineering is a structured discipline that helps programs and projects conceive, develop, integrate, test and deliver products and services that meet customer requirements within cost and schedule. This paper discusses the impact of Systems Engineering and a spiral development model on the success of the TOPAZ International Program and how the application of a similar approach could help ensure the success of future space reactor development projects.

Rover/NERVA-derived near-term nuclear propulsion

NASA Technical Reports Server (NTRS)

1993-01-01

FY-92 accomplishments centered on conceptual design and analyses for 25, 50, and 75 K engines with emphasis on the 50 K engine. During the first period of performance, flow and energy balances were prepared for each of these configurations and thrust-to-weight values were estimated. A review of fuel technology and key data from the Rover/NERVA program established a baseline for proven reactor performance and areas of enhancement to meet near-term goals. Studies were performed of the criticality and temperature profiles for probable fuel and moderator loadings for the three engine sizes, with a more detailed analysis of the 50 K size. During the second period of performance, analyses of the 50 K engine continued. A chamber/nozzle contour was selected and heat transfer and fatigue analyses were performed for likely construction materials. Reactor analyses were performed to determine component radiation heating rates, reactor radiation fields, water immersion poisoning requirements, temperature limits for restartability, and a tie-tube thermal analysis. Finally, a brief assessment of key enabling technologies was made, with a view toward identifying development issues and identification of the critical path toward achieving engine qualification within 10 years.

ETR COOLING TOWER. PUMP HOUSE (TRA645) IN SHADOW OF TOWER ...

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

ETR COOLING TOWER. PUMP HOUSE (TRA-645) IN SHADOW OF TOWER ON LEFT. AT LEFT OF VIEW, HIGH-BAY BUILDING IS ETR. ONE STORY ATTACHMENT IS ETR ELECTRICAL BUILDING. STACK AT RIGHT IS ETR STACK; MTR STACK IS TOWARD LEFT. CAMERA FACING NORTHEAST. INL NEGATIVE NO. 56-3799. Jack L. Anderson, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO ...

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

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO AN MTR BEAM HOLE. ONE MAN CHECKS RADIATION LEVEL AT THE END OF THE UNIVERSAL COFFIN, WHILE ANOTHER USES TOOL TO INSERT PLUG INTO HOLE THROUGH COFFIN. MEN WEAR "ANTI-C" (ANTI-CONTAMINATION) CLOTHING. INL NEGATIVE NO. 6198. R.G. Larsen, Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642 AND TRA647. FLOOR PLANS FOR FIRST AND SECOND ...

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

ETR, TRA-642 AND TRA-647. FLOOR PLANS FOR FIRST AND SECOND FLOORS OF THE OFFICE AND CONTROL BUILDING ALONG THE NORTH WALL OF THE ETR BUILDING. HEALTH PHYSICS, OPERATIONS, AND CONTROL ROOM. AIRLOCK DOOR. OFFICES. STAIRWAY LOCATIONS. KAISER ETR-5528-MTR-642-A-3, 10/1955. INL INDEX NO. 532-0642-00-100911, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

VERAView

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

Lee, Ronald W.; Collins, Benjamin S.; Godfrey, Andrew T.

2016-12-09

In order to support engineering analysis of Virtual Environment for Reactor Analysis (VERA) model results, the Consortium for Advanced Simulation of Light Water Reactors (CASL) needs a tool that provides visualizations of HDF5 files that adhere to the VERAOUT specification. VERAView provides an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The Python-based software provides instantaneous 2D and 3D images, 1D plots, and alphanumeric data from VERA multi-physics simulations.

MTR WING, TRA604. PRECAST CONCRETE PANELS AND DIMENSIONS. TYPES A, ...

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

MTR WING, TRA-604. PRECAST CONCRETE PANELS AND DIMENSIONS. TYPES A, B, C, D, E, AND F; AND HOW THEY ARE CONNECTED. TYPES C AND D ARE ON WEST SIDE WHERE GLASS BLOCKS SURROUND ENTRY DOOR. BLAW-KNOX 3150-804-20, SHEET #1, 11/1950. INL INDEX NO. 531-0604-62-098-100644, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. CAMERA FACES SOUTH AND LOOKS ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. CAMERA FACES SOUTH AND LOOKS AT DOOR TO M-3 CUBICLE. CUBICLE WALLS ARE MADE OF LEAD SHIELDING BRICKS. VALVE HANDLES AND STEMS PERTAIN TO SAMPLING. METAL SHIELDING DOOR. NOTE GLOVE BOX TO RIGHT OF CUBICLE DOOR. INL NEGATIVE NO. HD-46-21-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR HEAT EXCHANGER BUILDING, TRA644. SOUTH SIDE. CAMERA FACING NORTH. ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. SOUTH SIDE. CAMERA FACING NORTH. NOTE POURED CONCRETE WALLS. ETR IS AT LEFT OF VIEW. NOTE DRIVEWAY INSET AT RIGHT FORMED BY DEMINERALIZER WING AT RIGHT. SOUTHEAST CORNER OF ETR, TRA-642, IN VIEW AT UPPER LEFT. INL NEGATIVE NO. HD46-36-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

HOT CELL BUILDING, TRA632. FLOOR PLAN OF EXPANSION SHOWS LOCATION ...

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

HOT CELL BUILDING, TRA-632. FLOOR PLAN OF EXPANSION SHOWS LOCATION OF NEW CELLS, "HEAVY" CELL AT WEST END, "LIGHT" CELLS AT EAST. MOCK-UP AND STORAGE AREAS IN SOUTH HALF OF FLOOR. H.K. FERGUSON 895-MTR-ETR-632-A1, 12/1958. INL INDEX NO. 531-0632-00-279-101924, REV. 4. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLOOR AND ROOF PLANS FOR SECOND ...

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

PROCESS WATER BUILDING, TRA-605. FLOOR AND ROOF PLANS FOR SECOND FLOOR. DETAILS OF CONCRETE ROOF SLABS. FLASH EVAPORATOR SUPPORTS AND PIPE OPENINGS TO TANKS BELOW. NOTE SPECIFIES THAT EQUIPMENT IS TO BE INSTALLED BEFORE ERECTION OF ROOF AND WALLS. BLAW-KNOX 3150-805-4, 1/1951. INL INDEX NO. 531-0605-62-098-100660, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLOOR PLAN AND SECTION OF FLASH ...

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

PROCESS WATER BUILDING, TRA-605. FLOOR PLAN AND SECTION OF FLASH EVAPORATOR ROOM SHOWING ITS LOCATION ABOVE THE SEAL AND SUMP TANKS. PIPING TAKES WATER FROM SEAL TANK UPWARD TO FLASH EVAPORATORS AND THEN BACK DOWN TO SUMP TANK. BLAW-KNOX 3150-5-6, 8/1950. INL INDEX NO. 531-605-00-098-100011, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. BASEMENT. CUBICLE SHOWN IN ID33G101, ANOTHER ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. CUBICLE SHOWN IN ID-33-G-101, ANOTHER VIEW. PERSONNEL DOORWAY INTO CHAMBER IDENTIFIES SODIUM HAZARD AND POSSIBILITY OF INERT GAS. LIQUID SODIUM COOLANT WAS USED IN A SPECIAL ETR LOOP ADAPTED FOR IT IN 1972. INL NEGATIVE NO. HD24-3-2. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, NORTH HALF. CAMERA IS ...

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

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, NORTH HALF. CAMERA IS NEAR NORTHWEST CORNER AND FACING SOUTH ALONG WEST CORRIDOR. STORAGE CANAL IS ALONG LEFT OF VIEW; PERIMETER WALL, ALONG RIGHT. CORRIDOR WAS ONE MEANS OF WALKING FROM NORTH TO SOUTH SIDE OF CONSOLE FLOOR. INL NEGATIVE NO. HD46-18-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

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

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

1960-12-15

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

Apparatus for afterburning the exhaust gases of an internal combustion engine to remove pollutants therefrom

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

Laurent, P.A.

1976-09-28

An apparatus is described for afterburning the combustible pollutants from the exhaust gases of an internal combustion engine in a reactor, in which secondary air is introduced. Upstream of the reactor, a chamber in the form of a torus is provided, through which the exhaust gases from a maximum number of cylinders flow before entering the reactor. A first obstacle, acting as a flame holder is disposed inside the torus. The reactor comprises a chamber whose inner surface is approximately a surface of revolution, and mounted inside of which is a second obstacle, acting as flame holder, substantially along themore » axis of revolution. The second flame holder has a diameter large enough to provide a contact time of 1 to 3 x 10/sup -3/ seconds of the gas flow in a recirculation zone surrounding the second flame holder, the diameter of the second flame holder being 15 to 40 percent of the reactor diameter.« less

Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

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

Hatayama, Ariyoshi; Ogasawara, Masatada; Yamauchi, Michinori

1994-08-01

Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric)more » hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of I{sub p} = 15 MA, and toroidal field on the axis of B{sub o} = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is {approximately}7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs.« less

Low-Cost, Net-Shape Ceramic Radial Turbine Program

DTIC Science & Technology

1985-05-01

PROGRAM ELEMENT. PROJECT. TASK Garrett Turbine Engine Company AE OKUI UBR 111 South 34th Street, P.O. Box 2517 Phoenix, Arizona 85010 %I. CONTROLLING...processing iterations. Program management and materials characterization were conducted at Garrett Turbine Engine Company (GTEC), test bar and rotor...automotive gas turbine engine rotor development efforts at ACC. xvii PREFACE This is the final technical report of the Low-Cost, Net- Shape Ceramic

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

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

Kim, Seokho H; Berry, Jan

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclearmore » pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.« less