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Sample records for reactor joyo level

  1. Enhancement of Irradiation Capability of the Experimental Fast Reactor Joyo

    Science.gov (United States)

    Maeda, Shigetaka; Serine, Takashi; Aoyama, Takafumi; Suzuki, Soju

    2009-08-01

    The experimental fast reactor Joyo is the first sodium-cooled fast reactor in Japan. One of its primary missions is to perform irradiation tests of fuel and structural materials to support the development of fast reactors. The MK-III high performance core upgrade to enhance the irradiation testing capabilities was completed in 2003. In order to expand Joyo's capabilities for innovative irradiation testing applications, neutron spectrum tailoring, lower irradiation temperature, movable sample devices and fast neutron beam holes are being considered. This program responds to existing irradiation needs and aims to further expand capabilities for a variety of irradiation tests.

  2. A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO

    CERN Document Server

    Furuta, H; Hara, T; Haruna, T; Ishihara, N; Ishitsuka, M; Ito, C; Katsumata, M; Kawasaki, T; Konno, T; Kuze, M; Maeda, J; Matsubara, T; Miyata, H; Nagasaka, Y; Nitta, K; Sakamoto, Y; Suekane, F; Sumiyoshi, T; Tabata, H; Takamatsu, M; Tamura, N

    2011-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  3. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  4. Development of Observation Techniques in Reactor Vessel of Experimental Fast Reactor Joyo

    Science.gov (United States)

    Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

    In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

  5. Measurement and evaluation of Corrosion Products deposition distribution in the Experimental Fast Reactor JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Takafumi; Sumino, Kozo [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Masui, Tomohiko; Saikawa, Takuya

    1997-12-01

    The Corrosion Product (CP) is the major radiation source in the primary cooling system of an LMFBR plant. It is important to characterize and predict the CP behavior to reduce the personnel exposure dose due to CP deposition. The CP measurement was carried out in the Experimental Fast Reactor JOYO during the 11th annual inspection period when the accumulated reactor thermal power reached about 143 GWd. The CP deposition density was measured using a pure germanium detector. The plastic scintillation fiber (PSF) was applied for the gamma-ray dose rate distribution measurement and compared with the thermoluminescence dosimeter (TLD). The major results obtained by the CP measurements in JOYO are the follows: (1) The major CP nuclides deposited in the primary cooling system are {sup 54}Mn and {sup 60}Co. {sup 54}Mn is the dominant isotope and it tends to deposit in the cold leg region. On the other hand, {sup 60}Co deposits mainly in the hot leg region. The deposition density of {sup 54}Mn is about seven times as much as that of {sup 60}Co in the cold leg region and twice in the hot leg region. (2) The deposition densities of {sup 54}Mn and {sup 60}Co, and the gamma-dose rate were decreased from the last data in the previous annual inspection period mainly due to the short operation time and the longer cooling time. (3) The continuous gamma-ray dose rate distribution up to 10m can be measured by using the PSF in a few minutes. The PSF is suitable to measure the gamma-ray dose rate distribution in the maintenance work area where it is narrow and the mixture of gamma-ray sources from primary pipings and components. The data base of detailed gamma-ray dose rate distribution was greatly extended by the PSF. (author)

  6. Numerical analysis of irradiated Am samples in experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Sagara, Hiroshi; Yamamoto, Tetsuro; Shiba, Tomo-oki; Saito, Masaki [Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo, 1528550 (Japan); Koyama, Shin-ichi; Maeda, Shigetaka, E-mail: sagara@nr.titech.ac.jp [Japan Atomic Energy Agency, 4002 Nanta-cho, O-arai machi, Ibaraki, 3111393 (Japan)

    2010-03-15

    Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper deals with the numerical analysis of the Am sample irradiation in Joyo to examine the transmutation performance of pure isotope in fast neutron environment during the irradiation, and deals with the comparison with the experimental result to evaluate the accuracy of current available numerical tool. In {sup 241}Am pure isotope sample, the burn-up calculation of Am transmutation ratio and principal nuclides accumulation are agreed with the measured data within 1-{sigma} uncertainty caused of cross-section covariance. Isomeric ratio of {sup 242}Am in total {sup 241}Am capture reaction were calculated as 0.852{+-}0.016 in the core and 0.85{+-}0.025 in the axial and radial reactors. The current data and recently reported data by Koyama et. al 2008 support the latest version of nuclear data sets in ENDFB-VII and JENDL/AC-2008. From the view point of proliferation resistance, it was confirmed {sup 241}Amp reduces un-attractive Pu to abuse from the beginning to the end of irradiation, and it would have important role to denature Pu in future FBR based nuclear fuel cycle.

  7. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B{sub 4}C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between B{sub 4}C and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between B{sub 4}C and cladding, without heating B{sub 4}C or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1) The specification of a sodium bonded type control rod was determined with the wide gap between B{sub 4}C and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from B{sub 4}C in the sodium gap, was not a serious problem in the analysis which was considered. (2) A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3) The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4) The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed

  8. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    Energy Technology Data Exchange (ETDEWEB)

    G. Borges

    2006-01-31

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

  9. Verification of the Mimir-N2 Joyo plant dynamic code

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Akihiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kuroha, Takaya [Nuclear Energy System Inc., Tokyo (Japan)

    2002-06-01

    Passive safety systems at Experimental Fast Reactor JOYO were studied to demonstrate the inherent safety of MOX fueled sodium cooled fast breeder reactors. The Mimir-N2 analysis code, developed to analyze JOYO plant kinetics, was selected as the standard code. To increase the reliability, Mimir-N2 was modified based on data from plant characteristics and natural circulation tests in JOYO. JOYO operational data suggest that the burn-up dependency of the power reactivity coefficient could be due to the reactivity shift caused by decrease of fuel pellet thermal expansion in the axial direction. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, burn-up dependency was estimated and introduced to the Mimir-N2 model. This brought good correspondence between calculated and measured values for a step reactivity response test. Calculated plant parameters including power range neutron monitor response and fuel subassembly outlet coolant temperature corresponded to measured values. Mimir-N2 could simulate plant dynamics such as the perturbations due to core support plate thermal expansion. (author)

  10. Processing of Refractory Metal Alloys for JOYO Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    RF Luther; ME Petrichek

    2006-02-21

    This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang.

  11. Validation of intermediate heat and decay heat exchanger model in MARS-LMR with STELLA-1 and JOYO tests

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok; Hong, Jonggan; Yeom, Sujin; Eoh, Jaehyuk [Sodium-cooled Fast Reactor Design Division, Korea Atomic Energy Research Institute (KAERI), 989-111, Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Jeong, Hae-yong, E-mail: hyjeong@sejong.ac.kr [Department of Nuclear Engineering, Sejong University, 209 Neungdong-ro, Gwangjin-gu, Seoul 143-747 (Korea, Republic of)

    2016-11-15

    Highlights: • The capability of the MARS-LMR for heat transfer through IHX and DHX is evaluated. • Prediction of heat transfer through IHXs and DHXs is essential in the SFR analysis. • Data obtained from the STELLA-1 and the JOYO test are analyzed with the MARS-LMR. • MARS-LMR adopts the Aoki’s correlation for tube side and Graber-Rieger’s for shell. • The performance of the basic models and other available correlations is evaluated. • The current models in MARS-LMR show best prediction for JOYO and STELLA-1 data. - Abstract: The MARS-LMR code has been developed by the Korea Atomic Energy Research Institute (KAERI) to analyze transients in a pool-type sodium-cooled fast reactor (SFR). Currently, KAERI is developing a prototype Gen-IV SFR (PGSFR) with metallic fuel. The decay heat exchangers (DHXs) and the intermediate heat exchangers (IHXs) were designed as a sodium-sodium counter-flow tube bundle type for decay heat removal system (DHRS) and intermediate heat transport system (IHTS), respectively. The IHX and DHX are important components for a heat removal function under normal and accident conditions, respectively. Therefore, sodium heat transfer models for the DHX and IHX heat exchangers were added in MARS-LMR. In order to validate the newly added heat transfer model, experimental data were obtained from the JOYO and STELLA-1 facilities were analyzed. JOYO has two different types of IHXs: type-A (co-axial circular arrangement) and type-B (triangular arrangement). For the code validation, 38 and 39 data points for type A and type B were selected, respectively. A DHX performance test was conducted in STELLA-1, which is the test facility for heat exchangers and primary pump in the PGSFR. The DHX test in STELLA-1 provided eight data points for a code validation. Ten nodes are used in the heat transfer region is used, based on the verification test for the heat transfer models. RMS errors for JOYO IHX type A and type B of 19.1% and 4.3% are obtained

  12. Closeout of JOYO-1 Specimen Fabrication Efforts

    Energy Technology Data Exchange (ETDEWEB)

    ME Petrichek; JL Bump; RF Luther

    2005-10-31

    Fabrication was well under way for the JOYO biaxial creep and tensile specimens when the NR Space program was canceled. Tubes of FS-85, ASTAR-811C, and T-111 for biaxial creep specimens had been drawn at True Tube (Paso Robles, CA), while tubes of Mo-47.5 Re were being drawn at Rhenium Alloys (Cleveland, OH). The Mo-47.5 Re tubes are now approximately 95% complete. Their fabrication and the quantities produced will be documented at a later date. End cap material for FS-85, ASTAR-811C, and T-111 had been swaged at Pittsburgh Materials Technology, Inc. (PMTI) (Large, PA) and machined at Vangura (Clairton, PA). Cutting of tubes, pickling, annealing, and laser engraving were in process at PMTI. Several biaxial creep specimen sets of FS-85, ASTAR-811C, and T-111 had already been sent to Pacific Northwest National Laboratory (PNNL) for weld development. In addition, tensile specimens of FS-85, ASTAR-811C, T-111, and Mo-47.5 Re had been machined at Kin-Tech (North Huntington, PA). Actual machining of the other specimen types had not been initiated. Flowcharts 1-3 detail the major processing steps each piece of material has experienced. A more detailed description of processing will be provided in a separate document [B-MT(SRME)-51]. Table 1 lists the in-process materials and finished specimens. Also included are current metallurgical condition of these materials and specimens. The available chemical analyses for these alloys at various points in the process are provided in Table 2.

  13. Void swelling in high purity FeCrNi and FeCrNiTi alloys irradiated in JOYO

    Science.gov (United States)

    Muroga, T.; Araki, K.; Miyamoto, Y.; Yoshida, N.

    1988-07-01

    Microstructures have been observed in Fe-13Cr-14Ni and Fe-13Cr-14Ni-0.12Ti alloys irradiated in JOYO (Japanese Fast Experimental Reactor) at 400, 500 and 600 °C to the fluence of 0.079, 0.81 and 6.2 × 10 25n/ m2 ( E > 0.1 MeV). In the Fe-13Cr-14Ni alloy, voids are observed in all cases. The dose dependence of swelling seems to obey the kinetics of linear increase with or without initial short transient. On the other hand, remarkable swelling suppression effects are observed in the Fe-13Cr-14Ni-0.12Ti alloy. The detailed microstructural observation suggests the titanium addition effects suppress the void nucleation in the matrix by gettering impurities and obstructing dislocation climb by precipitate decoration on dislocation lines.

  14. Superior Charpy impact properties of ODS ferritic steel irradiated in JOYO

    Science.gov (United States)

    Kuwabara, T.; Kurishita, H.; Ukai, S.; Narui, M.; Mizuta, S.; Yamazaki, M.; Kayano, H.

    1998-10-01

    The effect of neutron irradiation on Charpy impact properties of an ODS ferritic steel developed by PNC was studied. The miniaturized Charpy V-notch (MCVN) specimens (1.5 × 1.5 × 20 mm) of two orientations (longitudinal, called 1DS-L, and transverse, 1DS-T) were irradiated to fluence levels of (0.3-3.8) × 10 26 n/m 2 ( E n > 0.1 MeV) between 646 and 845 K in JOYO. MCVN specimens before and after the irradiation were subjected to instrumented Charpy impact tests. The test results and fracture surface observations showed that in the unirradiated state the steel showed no ductile-to-brittle transition behavior until 153 K regardless of orientation and the upper shelf energy of the steel was as high as that of a high-strength ferritic steel without dispersed oxide. Such excellent impact properties were essentially maintained after the irradiation although an appreciable decrease in absorbed energy occurred by higher temperature irradiations at and above 793 K.

  15. Re-evaluation of seismic design for JOYO buildings and equipments

    Energy Technology Data Exchange (ETDEWEB)

    Isozaki, K.; Tomita, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-07-01

    Hyougo-ken southern earthquake broke out in 1997/01/17. The Atomic Energy Safety Commission considered reasonable of the design guide for seismic design. And the Science and Technology Agency (STA) required reevaluation of atomic power facilities built by old design guide according to the new seismic design guide. JOYO obtained the construction license in 1970/02. Heat transport system and buildings of JOYO was re-evaluated by the new seismic design guide for the MK-III project. So, JOYO was not required re-evaluation by STA. But, this evaluation of MK-III was limited to reconstruction area, and the seismic design was reevaluated extensively to confirm earthquake proof characteristics. The structural integrity of buildings and equipments was confirmed by the result of reevaluation by the new seismic design guide. The analysis model conditions were established according to the 1987 and 1991 version of JEAG. This was done by ground investigation result and buildings vibration test. It was made clear that the analysis model conditions were reasonable and conservative from a technical view point. (author)

  16. Burning high-level TRU waste in fusion fission reactors

    National Research Council Canada - National Science Library

    Shen, Yaosong

    2016-01-01

    .... A new method of burning high-level transuranic (TRU) waste combined with Thorium–Uranium (Th–U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper...

  17. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  18. Design of megawatt power level heat pipe reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-12

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

  19. Gamma thermometer based reactor core liquid level detector

    Science.gov (United States)

    Burns, Thomas J.

    1983-01-01

    A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

  20. Burning high-level TRU waste in fusion fission reactors

    Science.gov (United States)

    Shen, Yaosong

    2016-09-01

    Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

  1. Derivation methods for clearance levels applied to reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okoshi, Minoru; Seki, Takeo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    In order to support the discussion by the Nuclear Safety Commission, JAERI derived the unconditional clearance levels for concrete and metal arising from the operation and dismantling of nuclear reactors. The clearance levels of 20 radionuclides were derived from 10 {mu}Sv/y of individual doses by deterministic approach. In this approach, calculation models were established to assess individual doses resulting from 73 exposure pathways related to disposal and recycle/reuse, and realistic parameter values were selected considering Japanese natural and social conditions. The appropriateness of selected parameter values was confirmed by stochastic analyses. (author)

  2. Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-06-15

    Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

  3. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  4. Level 1 Tornado PRA for the High Flux Beam Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

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

    Science.gov (United States)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  6. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    Science.gov (United States)

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  7. Calibration of a He accumulation fluence monitor for fast reactor dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Chikara [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-03-01

    The helium accumulation fluence monitor (HAFM) has been developed for a fast reactor dosimetry. The HAFM measurement system was calibrated using He gas and He implanted samples and the measurement accuracy was confirmed to be less than 5%. Based on the preliminary irradiation test in JOYO, the measured He in the {sup 10}B type HAFM agreed well with the calculated values using the JENDL-3.2 library. (author)

  8. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

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

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

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

  10. Nuclear reactor high-level waste: origin and safe disposal

    Energy Technology Data Exchange (ETDEWEB)

    Chua, C.; Tsipis, K. (Massachusetts Inst. of Tech., Cambridge, MA (USA))

    High-level waste (HLW) is a natural component of the nuclear fuel cycle. Because of its radioactivity, HLW needs to be handled with great care. Different alternatives for permanently storing HLW are evaluated. Studies have shown that the disposal of HLW is safest when the waste is first vitrified before storage. Simple calculations show that vitrified HLW that is properly buried in deep, carefully chosen crystalline rock structures poses insignificant health risks. (author).

  11. Bioaugmentation of syntrophic acetate-oxidizing culture in biogas reactors exposed to increasing levels of ammonia.

    Science.gov (United States)

    Westerholm, Maria; Levén, Lotta; Schnürer, Anna

    2012-11-01

    The importance of syntrophic acetate oxidation for process stability in methanogenic systems operating at high ammonia concentrations has previously been emphasized. In this study we investigated bioaugmentation of syntrophic acetate-oxidizing (SAO) cultures as a possible method for decreasing the adaptation period of biogas reactors operating at gradually increased ammonia concentrations (1.5 to 11 g NH(4)(+)-N/liter). Whole stillage and cattle manure were codigested semicontinuously for about 460 days in four mesophilic anaerobic laboratory-scale reactors, and a fixed volume of SAO culture was added daily to two of the reactors. Reactor performance was evaluated in terms of biogas productivity, methane content, pH, alkalinity, and volatile fatty acid (VFA) content. The decomposition pathway of acetate was analyzed by isotopic tracer experiments, and population dynamics were monitored by quantitative PCR analyses. A shift in dominance from aceticlastic methanogenesis to SAO occurred simultaneously in all reactors, indicating no influence by bioaugmentation on the prevailing pathway. Higher abundances of Clostridium ultunense and Tepidanaerobacter acetatoxydans were associated with bioaugmentation, but no influence on Syntrophaceticus schinkii or the methanogenic population was distinguished. Overloading or accumulation of VFA did not cause notable dynamic effects on the population. Instead, the ammonia concentration had a substantial impact on the abundance level of the microorganisms surveyed. The addition of SAO culture did not affect process performance or stability against ammonia inhibition, and all four reactors deteriorated at high ammonia concentrations. Consequently, these findings further demonstrate the strong influence of ammonia on the methane-producing consortia and on the representative methanization pathway in mesophilic biogas reactors.

  12. Evaluation of strategies for end storage of high-level reactor fuel; Vurdering av strategier for sluttlagring av hoeyaktivt reaktorbrensel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized.

  13. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  14. Nonlinear Adaptive Dynamic Output-Feedback Power-Level Control of Nuclear Heating Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2013-01-01

    Full Text Available Due to the high safety performance of small nuclear reactors, there is a promising future for small reactors. Nuclear heating reactor (NHR is a small reactor that has many advanced safety features such as the integrated arrangement, natural circulation at any power levels, self-pressurization, hydraulic control rod driving, and passive residual heating removing and can be applied to the fields of district heating, seawater desalination, and electricity production. Since the NHR dynamics has strong nonlinearity and uncertainty, it is meaningful to develop the nonlinear adaptive power-level control technique. From the idea of physically based control design method, a novel nonlinear adaptive power-level control is given for the NHR in this paper. It is theoretically proved that this newly built controller does not only provide globally asymptotic closed-loop stability but is also adaptive to the system uncertainty. Numerical simulation results show the feasibility of this controller and the relationship between the performance and controller parameters.

  15. Irradiated Xenon Isotopic Ratio Measurement for Failed Fuel Detection and Location in Fast Reactor

    Science.gov (United States)

    Ito, Chikara; Iguchi, Tetsuo; Harano, Hideki

    2009-08-01

    The accuracy of xenon isotopic ratio burn-up calculations used for failed fuel identification was evaluated by an irradiation test of xenon tag gas samples in the Joyo test reactor. The experiment was carried out using pressurized steel capsules containing unique blend ratios of stable xenon tag gases in an on-line creep rupture experiment in Joyo. The tag gas samples were irradiated to total neutron fluences of 1.6 to 4.8 × 1026 n/m2. Laser resonance ionization mass spectrometry was used to analyze the cover gas containing released tag gas diluted to isotopic ratios of 100 to 102 ppb. The isotopic ratios of xenon tag gases after irradiation were calculated using the ORIGEN2 code. The neutron cross sections of xenon nuclides were based on the JENDL-3.3 library. These cross sections were collapsed into one group using the neutron spectra of Joyo. The comparison of measured and calculated xenon isotopic ratios provided C/E values that ranged from 0.92 to 1.10. The differences between calculation and measurement were considered to be mainly due to the measurement errors and the xenon nuclide cross section uncertainties.

  16. Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kurisaka, K., E-mail: kurisaka.kennichi@jaea.go.jp [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Sakai, T.; Yamano, H. [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Fujita, S.; Minagawa, K. [Department of Mechanical Engineering, School of Engineering, Tokyo Denki University, Tokyo (Japan); Yamaguchi, A.; Takata, T. [Department of Energy and Environment Engineering, Osaka University, Osaka (Japan)

    2014-04-01

    This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these

  17. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  18. Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-01-30

    The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, and at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.

  19. A novel two-level dielectric barrier discharge reactor for methyl orange degradation.

    Science.gov (United States)

    Tao, Xumei; Wang, Guowei; Huang, Liang; Ye, Qingguo; Xu, Dongyan

    2016-12-15

    A novel pilot two-level dielectric barrier discharge (DBD) reactor has been proposed and applied for degradation of continuous model wastewater. The two-level DBD reactor was skillfully realized with high space utilization efficiency and large contact area between plasma and wastewater. Various conditions such as applied voltage, initial concentration and initial pH value on methyl orange (MO) model wastewater degradation were investigated. The results showed that the appropriate applied voltage was 13.4 kV; low initial concentration and low initial pH value were conducive for MO degradation. The percentage removal of 4 L MO with concentration of 80 mg/L reached 94.1% after plasma treatment for 80min. Based on ultraviolet spectrum (UV), Infrared spectrum (IR), liquid chromatography-mass spectrometry (LC-MS) analysis of degradation intermediates and products, insights in the degradation pathway of MO were proposed.

  20. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  1. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yi-Ting [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Yu, Yi-Hui [Department of Civil Engineering, National Taiwan University, Taipei 106, Taiwan (China); Nguyen, Van-Huy [Department of Chemical Engineering, National Taiwan University of Science and Technology, Taipei 10617, Taiwan (China); Lu, Kung-Te [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Wu, Jeffrey Chi-Sheng, E-mail: cswu@ntu.edu.tw [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Chang, Luh-Maan [Department of Civil Engineering, National Taiwan University, Taipei 106, Taiwan (China); Kuo, Chi-Wen [Taiwan Semiconductor Manufacturing Company, Hsinchu 30078, Taiwan (China)

    2013-11-15

    Graphical abstract: We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. -- Highlights: • The combination of optical fiber and honeycomb significantly enhanced the performance of VOCs photodegradation. • The removal efficiency of m-xylene is enhanced to 96.5% as compared to 22.0% for UV irradiation alone. • Fiber-illuminated honeycomb reactor is the first step toward an industrial-scale technology on the removal of xylene. -- Abstract: The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO{sub 2} photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO{sub 2} selectivity. Interestingly, Mn-TiO{sub 2} in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future.

  2. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

  3. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji [Malaysian Nuclear Agency, MOSTI, Bangi, 43000 Kajang, Selangor (Malaysia); Brayon, Fedrick Charlie Matthew [Atomic Energy Licensing Board, MOSTI, 43800 Dengkil, Selangor (Malaysia); Mohamed, Faizal [Universiti Kebangsaan Malaysia, Bangi, Selangor (Malaysia)

    2014-02-12

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  4. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    Science.gov (United States)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji

    2014-02-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  5. The Effect of Duct Level on the Performance of Reactor Vault Cooling System in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Sujin; Ryu, Seung Ho; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Development of the prototype gen-Ⅵ sodium-cooled fast reactor (PGSFR) has been ongoing in Korea Atomic Energy Research Institute (KAERI). A reactor vault cooling system (RVCS), one of passive decay heat removal systems (PDHRS), passively removes core decay heat by chimney effect when severe accidents occur. The air cooling path is located around containment vessel (CV). An air separator which divides the downstream air and the upstream air is installed between CV and the concrete wall. To design the RVCS, key design parameters such as stack height, gap size between the concrete wall and the air separator, gap size between the air separator and the CV, thickness and layer composition of the air separator have to be determined. A duct level is one of these design parameters. It denotes the height of the upstream air path and related to the heat transfer length from CV to air. The duct level should be optimized with considering structural reliability and heat removal performance. Thus, in this paper, the heat removal performance of RVCS is evaluated depends on the duct level using 1D system design code, that is developed by KAERI autonomously, and commercial CFD program for optimum design of RVCS In this paper, the heat removal performance of RVCS is evaluated depends on the duct level using PARS2- LMR code and commercial CFD program for optimum design of RVCS to satisfy both conflicting needs, structural reliability and cooling performance. As a result of PARS2-LMR code analysis, it was observed that the heat removal rate increases as increase of duct level and the geometrical conditions, that satisfy the design limitations, were obtained. To qualitatively observe the trends of local temperature distribution, CFD simulations were conducted and hotspots were observed at the upper region of ducts for the low duct level case.

  6. Preliminary Requirement of Hot Pool Free Surface Level from PGSFR Reactor Head

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeonghoi; Joo, Hyeongkook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sensitivity study on structural integrity evaluations are carried out to make a decision of a hot pool free surface location from the reactor head for a preliminary designed reactor enclosure system. To do this, the thermal stress evaluations for a reactor vessel are carried out for a steady state normal operating condition with detailed heat transfer analyses through the reactor enclosure system. From these results, the preliminary design requirement of a hot pool free surface location from the reactor head is established to be 2.0m. From the sensitivity studies on the structural integrity evaluations for a steady state condition, the preliminary distance from the hot pool free surface to the reactor head is determined to be 2.0m same as a conceptual design. More detailed structural analyses for a reactor enclosure system will be carried out as a PGSFR structural design goes forward in detail.

  7. Main technical options of the Jules Horowitz reactor project to achieve high flux performances and high safety level

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 kW/I for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: - A safety approach based upon the defence-in-depth principle. - A strategy of generic approaches to assess experimental risks in the facility. - Internal events analysis taking into account risks linked to reactor and experiments (e.g., radioactive source-term). - Systematic consideration of external hazards (e.g., earthquake, airplane crash) and internal hazards. - Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (authors)

  8. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level.

    Science.gov (United States)

    Wu, Yi-Ting; Yu, Yi-Hui; Nguyen, Van-Huy; Lu, Kung-Te; Wu, Jeffrey Chi-Sheng; Chang, Luh-Maan; Kuo, Chi-Wen

    2013-11-15

    The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO2 photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO2 selectivity. Interestingly, Mn-TiO2 in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  11. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

  12. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  13. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  14. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  15. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, M.L.; Gibson, L.T. [Oak Ridge National Laboratory, TN (United States); Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  16. Submersible microbial desalination cell for simultaneous ammonia recovery and electricity production from anaerobic reactors containing high levels of ammonia

    DEFF Research Database (Denmark)

    Zhang, Yifeng; Angelidaki, Irini

    2015-01-01

    High ammonia concentration in anaerobic reactors can seriously inhibit the anaerobic digestion process. In this study, a submersible microbial desalination cell (SMDC) was developed as an innovative method to lower the ammonia level in a continuous stirred tank reactor (CSTR) by in situ ammonia...... recovery and electricity production. In batch experiment, the ammonia concentration in the CSTR decreased from 6 to 0.7g-N/L during 30days, resulting in an average recovery rate of 80g-N/m2/d. Meanwhile, a maximum power density of 0.71±0.5W/m2 was generated at 2.85A/m2. Both current driven NH4+ migration...

  17. Steps towards verification and validation of the Fetch code for Level 2 analysis, design, and optimization of aqueous homogeneous reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nygaard, E. T. [Babcock and Wilcox Technical Services Group, 800 Main Street, Lynchburg, VA 24504 (United States); Pain, C. C.; Eaton, M. D.; Gomes, J. L. M. A.; Goddard, A. J. H.; Gorman, G.; Tollit, B.; Buchan, A. G.; Cooling, C. M. [Applied Modelling and Computation Group, Dept. of Earth Science and Engineering, Imperial College London, SW7 2AZ (United Kingdom); Angelo, P. L. [Y-12 National Security Complex, Oak Ridge, TN 37831 (United States)

    2012-07-01

    Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' While AHRs have been modeled effectively using simplified 'Level 1' tools, the complex interactions between fluids, neutronics, and solid structures are important (but not necessarily safety significant). These interactions require a 'Level 2' modeling tool. Imperial College London (ICL) has developed such a tool: Finite Element Transient Criticality (FETCH). FETCH couples the radiation transport code EVENT with the computational fluid dynamics code (Fluidity), the result is a code capable of modeling sub-critical, critical, and super-critical solutions in both two-and three-dimensions. Using FETCH, ICL researchers and B and W engineers have studied many fissioning solution systems include the Tokaimura criticality accident, the Y12 accident, SILENE, TRACY, and SUPO. These modeling efforts will ultimately be incorporated into FETCH'S extensive automated verification and validation (V and V) test suite expanding FETCH'S area of applicability to include all relevant physics associated with AHRs. These efforts parallel B and W's engineering effort to design and optimize an AHR to produce Mo99. (authors)

  18. A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

    Science.gov (United States)

    Menke, M. M.; Judd, B. R.

    1973-01-01

    The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

  19. Design considerations regarding slug ruptures in the intermediate power level reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pearl, W.L.; Pursel, C.A.

    1954-11-01

    The minimum shutdown time, to permit accessibility, for the Intermediate Power Reactor is estimated to be 38 hours. In case the reactor were shutdown following each rupture this long shutdown period would have serious disadvantages. The desirability of being able to make firm power commitments (independent of slug ruptures) has led to a study of the possibility of continuous operation following a rupture. There is evidence to indicate that, at the proposed water temperature, the rate of corrosion of uranium may be so high that at least a major portion of the rupture products may have entered the system before the reactor can be shutdown. A pushout of the affected column would then be a pushout of only those slugs which are still intact and the problem would still remain of removing the rupture products from the system. The first portion of this report is concerned with the rate of corrosion of a slug following rupture and the possible limitations to the principle of non-shutdown operation. These limitations include a flow stoppage by the ruptured can, undue increase in gamma activity, increased corrosion by the rupture products, and adherence of rupture products to the piping. The latter portion of the document is concerned with design considerations of the shielding and water plant so as to eliminate or minimize the effects of the introduction of rupture products into the cooling system. 7 refs., 2 figs.

  20. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  1. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

  2. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  3. Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan)

    2002-12-01

    In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

  4. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    Science.gov (United States)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  5. Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2015-11-15

    Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

  6. Annual report of Power Reactor and Nuclear Fuel Development Corporation, fiscal year 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This was the Annual Report of the Power Reactor and Nuclear Fuel Development Corporation, Fiscal Year of 1994. In this report, the following 12 items are described: (1) Development of the fast breeding reactor; (a) operation of the fast experimental reactor, `Joyo`, (b) construction and trial operation of the fast breeding prototype reactor, `Monju`, and (c) R and D of FBR; (2) Development of the new type conversion reactor; (a) operation of prototype reactor, `Fugen`, and (b) R and D of ATR; (3) Development of uranium mining and conversion; (4) Development of uranium concentration technology; (5) Development of plutonium fuel; (a) preparation of the MOX fuel, (b) preparation facility construction of the MOX fuel, (c) R and D of plutonium fuel. and (d) technical development of plutonium mixing and conversion; (6) Reprocessing of spent fuel; (7) Environmental technology development of radioactive waste; (8) Creative and innovative R and D; (9) Management and nuclear non-proliferation countermeasure of nuclear matter; (10) Safety management and safety study; (11) Related common business; and (12) General management business. (G.K.)

  7. Submersible microbial desalination cell for simultaneous ammonia recovery and electricity production from anaerobic reactors containing high levels of ammonia.

    Science.gov (United States)

    Zhang, Yifeng; Angelidaki, Irini

    2015-02-01

    High ammonia concentration in anaerobic reactors can seriously inhibit the anaerobic digestion process. In this study, a submersible microbial desalination cell (SMDC) was developed as an innovative method to lower the ammonia level in a continuous stirred tank reactor (CSTR) by in situ ammonia recovery and electricity production. In batch experiment, the ammonia concentration in the CSTR decreased from 6 to 0.7 g-N/L during 30 days, resulting in an average recovery rate of 80 g-N/m(2)/d. Meanwhile, a maximum power density of 0.71±0.5 W/m(2) was generated at 2.85 A/m(2). Both current driven NH4(+) migration and free NH3 diffusion were identified as the mechanisms responsible for the ammonia transportation. With an increase in initial ammonia concentration and a decrease in external resistance, the SMDC performance was enhanced. In addition, the coexistence of other cations in CSTR or cathode had no negative effect on the ammonia transportation. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. Determination of the level of water in the core of reactors PWR using neutron detectors signal ex core; Determinacion del nivel del agua del nucleo de reactores PWR usando la senal de detectores neutronicos excore

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Miro, R.; Verdu, G.

    2014-07-01

    The level of water from the core provides relevant information of the neutronic and thermal hydraulic of the reactor as the power, k EFF and cooling capacity. In fact, this level monitoring can be used for prediction of LOCA and reduction of cooling that can cause damage to the core. There are several teams that measure a variety of parameters of the reactor, as opposed to the level of the water of the core. However, the detectors 'excore' measure fast neutrons which escape from the core and there are studies that demonstrate the existence of a relationship between them and the water level of the kernel due to the water shield. Therefore, a methodology has been developed to determine this relationship, using the Monte Carlo method using the MCNP code and apply variance reduction techniques based on the attached flow that is obtained using the method of discrete ordinates using code TORT. (Author)

  9. Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2011-11-01

    Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

  10. Modeling and monitoring cyclic and linear volatile methylsiloxanes in a wastewater treatment plant using constant water level sequencing batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, De-Gao, E-mail: degaowang@dlmu.edu.cn; Du, Juan; Pei, Wei; Liu, Yongjun; Guo, Mingxing

    2015-04-15

    The fate of cyclic and linear volatile methylsiloxanes (VMSs) was evaluated in a wastewater treatment plant (WWTP) using constant water level sequencing batch reactors from Dalian, China. Influent, effluent, and sewage sludge samples were collected for seven consecutive days. The mean concentrations of cyclic VMSs (cVMSs) in influent and effluent samples are 1.05 μg L{sup −1} and 0.343 μg L{sup −1}; the total removal efficiency of VMSs is > 60%. Linear VMS (lVMS) concentration is under the quantification limitation in aquatic samples but is found in sludge samples with a value of 90 μg kg{sup −1}. High solid-water partition coefficients result in high VMS concentrations in sludge with the mean value of 5030 μg kg{sup −1}. No significant differences of the daily mass flows are found when comparing the concentration during the weekend and during working days. The estimated mass load of total cVMSs is 194 mg d{sup −1} 1000 inhabitants{sup −1} derived for the population. A mass balance model of the WWTP was developed and derived to simulate the fate of cVMSs. The removal by sorption on sludge increases, and the volatilization decreases with increasing hydrophobicity and decreasing volatility for cVMSs. Sensitivity analysis shows that the total suspended solid concentration in the effluent, mixed liquor suspended solid concentration, the sewage sludge flow rate, and the influent flow rate are the most influential parameters on the mass distribution of cVMSs in this WWTP. - Highlights: • A mass balance model for siloxanes was developed in sequencing batch reactor. • Total suspended solid in effluent has the most influence on removal efficiency. • Enhancement of suspended solid removal reduces the release to aquatic environment.

  11. Air radioactivity levels following the Fukushima reactor accident measured at the Laboratoire Souterrain de Modane, France.

    Science.gov (United States)

    Loaiza, P; Brudanin, V; Piquemal, F; Reyss, J-L; Stekl, I; Warot, G; Zampaolo, M

    2012-12-01

    The radioactivity levels in the air of the radionuclides released by the Fukushima accident were measured at the Laboratoire Souterrain de Modane, in the South-East of France, during the period 25 March-18 April 2011. Air-filters from the ventilation system exposed for one or two days were measured using low-background gamma-ray spectrometry. In this paper we present the activity concentrations obtained for the radionuclides (131)I, (132)Te, (134)Cs, (137)Cs, (95)Nb, (95)Zr, (106)Ru, (140)Ba/La and (103)Ru. The activity concentration of (131)I was of the order of 100 μBq/m(3), more than 100 times higher than the activities of other fission products. The highest activities of (131)I were measured as a first peak on 30 March and a second peak on 3-4 April. The activity concentrations of (134)Cs and (137)Cs varied from 5 to 30 μBq/m(3). The highest activity concentration recorded for Cs corresponded to the same period as for (131)I, with a peak on 2-3 April. The results of the radioactivity concentration levels in grass and mushrooms exposed to the air in the Modane region were also measured. Activity concentrations of (131)I of about 100 mBq/m(2) were found in grass.

  12. A sequencing batch reactor system for high-level biological nitrogen and phosphorus removal from abattoir wastewater.

    Science.gov (United States)

    Lemaire, Romain; Yuan, Zhiguo; Bernet, Nicolas; Marcos, Marcelino; Yilmaz, Gulsum; Keller, Jürg

    2009-06-01

    A sequencing batch reactor (SBR) system is demonstrated to biologically remove nitrogen, phosphorus and chemical oxygen demand (COD) to very low levels from abattoir wastewater. Each 6 h cycle contained three anoxic/anaerobic and aerobic sub-cycles with wastewater fed at the beginning of each anoxic/anaerobic period. The step-feed strategy was applied to avoid high-level build-up of nitrate or nitrite during nitrification, and therefore to facilitate the creation of anaerobic conditions required for biological phosphorus removal. A high degree removal of total phosphorus (>98%), total nitrogen (>97%) and total COD (>95%) was consistently and reliably achieved after a 3-month start-up period. The concentrations of total phosphate and inorganic nitrogen in the effluent were consistently lower than 0.2 mg P l(-1) and 8 mg N l(-1), respectively. Fluorescence in situ hybridization revealed that the sludge was enriched in Accumulibacter spp. (20-40%), a known polyphosphate accumulating organism, whereas the known glycogen accumulating organisms were almost absent. The SBR received two streams of abattoir wastewater, namely the effluent from a full-scale anaerobic pond (75%) and the effluent from a lab-scale high-rate pre-fermentor (25%), both receiving raw abattoir wastewater as feed. The pond effluent contained approximately 250 mg N l(-1) total nitrogen and 40 mg P l(-1) of total phosphorus, but relatively low levels of soluble COD (around 500 mg l(-1)). The high-rate lab-scale pre-fermentor, operated at 37 degrees C and with a sludge retention time of 1 day, proved to be a cheap and effective method for providing supplementary volatile fatty acids allowing for high-degree of biological nutrient removal from abattoir wastewater.

  13. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  14. Contamination level of four priority phthalates in North Indian wastewater treatment plants and their fate in sequencing batch reactor systems.

    Science.gov (United States)

    Gani, Khalid Muzamil; Rajpal, Ankur; Kazmi, Absar Ahmad

    2016-03-01

    The contamination level of four phthalates in untreated and treated wastewater of fifteen wastewater treatment plants (WWTPs) and their fate in a full scale sequencing batch reactor (SBR) based WWTP was evaluated in this study. The four phthalates were diethyl phthalate (DEP), dibutyl phthalate (DBP), benzylbutyl phthalate (BBP) and diethylhexyl phthalate (DEHP). All compounds were present in untreated wastewater with DEHP being present in the highest mean concentration of 28.4 ± 5.3 μg L(-1). The concentration was in the range of 7.3 μg L(-1) (BBP) to 28.4 μg L(-1) (DEHP) in untreated wastewater and 1.3 μg L(-1) (DBP) to 2.6 μg L(-1) (DEHP) in treated wastewater. The nutrient removal process and advance tertiary treatment based WWTPs showed the highest phthalate removal efficiencies of 87% and 93%, respectively. The correlation between phthalate removal and conventional performance of WWTPs was positive. Fate analysis of these phthalates in a SBR based WWTP showed that total removal of the sum of phthalates in a primary settling tank and SBR was 84% out of which 55% is removed by biodegradation and 29% was removed by sorption to primary and secondary sludge. The percentage removal of four phthalates in primary settling tanks was 18%. Comparison of the diluted effluent DEHP concentration with its environmental quality standards showed that the dilution in an effluent receiving water body can reduce the DEHP emissions to acceptable values.

  15. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  16. Modeling and monitoring cyclic and linear volatile methylsiloxanes in a wastewater treatment plant using constant water level sequencing batch reactors.

    Science.gov (United States)

    Wang, De-Gao; Du, Juan; Pei, Wei; Liu, Yongjun; Guo, Mingxing

    2015-04-15

    The fate of cyclic and linear volatile methylsiloxanes (VMSs) was evaluated in a wastewater treatment plant (WWTP) using constant water level sequencing batch reactors from Dalian, China. Influent, effluent, and sewage sludge samples were collected for seven consecutive days. The mean concentrations of cyclic VMSs (cVMSs) in influent and effluent samples are 1.05 μg L(-1) and 0.343 μg L(-1); the total removal efficiency of VMSs is >60%. Linear VMS (lVMS) concentration is under the quantification limitation in aquatic samples but is found in sludge samples with a value of 90 μg kg(-1). High solid-water partition coefficients result in high VMS concentrations in sludge with the mean value of 5030 μg kg(-1). No significant differences of the daily mass flows are found when comparing the concentration during the weekend and during working days. The estimated mass load of total cVMSs is 194 mg d(-1)1000 inhabitants(-1) derived for the population. A mass balance model of the WWTP was developed and derived to simulate the fate of cVMSs. The removal by sorption on sludge increases, and the volatilization decreases with increasing hydrophobicity and decreasing volatility for cVMSs. Sensitivity analysis shows that the total suspended solid concentration in the effluent, mixed liquor suspended solid concentration, the sewage sludge flow rate, and the influent flow rate are the most influential parameters on the mass distribution of cVMSs in this WWTP.

  17. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  18. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  19. Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

    2015-06-19

    This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

  20. Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

    Energy Technology Data Exchange (ETDEWEB)

    Johanna Oxstrand; Katya Le Blanc

    2014-07-01

    The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

  1. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  2. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  3. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  4. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  5. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  6. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  7. Method to identify potential phosphorus rate-limiting conditions in post-denitrification biofilm reactors within systems designed for simultaneous low-level effluent nitrogen and phosphorus concentrations.

    Science.gov (United States)

    Boltz, Joshua P; Morgenroth, Eberhard; Daigger, Glen T; deBarbadillo, Christine; Murthy, Sudhir; Sørensen, Kim H; Stinson, Beverly

    2012-12-01

    Water-quality standards requiring simultaneous low level effluent N and P concentrations are increasingly common in Europe and the United States of America. Moving bed biofilm reactors (MBBRs) and biologically active filters (BAFs) have been used as post-denitrification biofilm reactors in processes designed and operated for this purpose (Boltz et al., 2010a). There is a paucity of information describing systematic design and operational protocols that will minimize the potential for phosphorus rate-limited conditions as well as a lack of information describing the interaction between these post-denitrification biofilm reactors and unit processes that substantially alter phosphorus speciation (e.g., chemically enhanced clarification). In this paper, a simple mathematical model for estimating the threshold below which P becomes rate-limiting, and the model is presented and evaluated by comparing its predictions with operational data from post-denitrification MBBRs and BAFs. Ortho-phosphorus (PO(4)-P), which is the dissolved reactive component of total phosphorus, was a primary indicator of P rate-limiting conditions in the evaluated post-denitrification biofilm reactors. The threshold below which PO(4)-P becomes the rate-limiting substrate is defined: S(PO4-P):S(NOx-N) = 0.0086 g P/g N and S(PO4-P):S(M) = 0.0013 g P/g COD. Additional analyses indicate J(NOx-N)(avg) =0.48 g/m2/d when S(PO4-P):S(NOx-N) > 0.0086, and J(NOx-N)(avg) = 0.06 g/m2/d when S(PO4-P):S(NOx-N) clarifiers effluent stream, dosing phosphoric acid in the MBBR or BAF influent stream, and/or optimizing secondary process EBPR may overcome phosphorus rate-limitations in the biofilm-based post-denitrification process. Copyright © 2012 Elsevier Ltd. All rights reserved.

  8. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of

  9. Membrane reactor. Membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))

    1990-08-05

    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  10. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    Science.gov (United States)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-09-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  11. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  12. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  13. Effect of the variation of the level of lactose conversion in an immobilized lactase reactor upon operating costs for the production of Baker's yeast from hydrolyzed permeate obtained from the ultrafiltration of cottage cheese whey

    Energy Technology Data Exchange (ETDEWEB)

    Scott, T.C.; Hill, C.G. Jr.; Amundson, C.H.

    1987-01-01

    Operating costs for the production of Baker's yeast from hydrolyzed permeate from the ultrafiltration of cottage cheese whey were calculated as a function of the level of lactose conversion in the immobilized lactase reactor. These costs were calculated for the case of 90% conversion of lactose in the reactor and compared to those which result when running the reactor at lower conversions with recycle of unreacted lactose. Total operating costs were estimated by combining individual operating costs for the immobilized enzyme reactor, costs associated with processing a lactose recycle stream, and energy costs associated with cooling the reactor feed stream and sterilizing the hydrolysate stream. It was determined that operating costs are minimized at about 9.9 cents per pound of lactose when the reactor is run at approximately 72% conversion. This represents a savings of 2.4 cents per pound of lactose over the case of a once-through 90% conversion of lactose in the reactor. 8 refs., 4 figs., 9 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

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

  15. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  16. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  17. Initiating Events for Multi-Reactor Plant Sites

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  18. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  19. Effect of nitrate concentration on filamentous bulking under low level of dissolved oxygen in an airlift inner circular anoxic-aerobic incorporate reactor.

    Science.gov (United States)

    Su, Yiming; Zhang, Yalei; Zhou, Xuefei; Jiang, Ming

    2013-09-01

    This laboratory research investigated a possible cause of filamentous bulking under low level of dissolved oxygen conditions (dissolved oxygen value in aerobic zone maintained between 0.6-0.8 mg O2/L) in an airlift inner-circular anoxic-aerobic reactor. During the operating period, it was observed that low nitrate concentrations affected sludge volume index significantly. Unlike the existing hypothesis, the batch tests indicated that filamentous bacteria (mainly Thiothrix sp.) could store nitrate temporarily under carbon restricted conditions. When nitrate concentration was below 4 mg/L, low levels of carbon substrates and dissolved oxygen in the aerobic zone stimulated the nitrate-storing capacity of filaments. When filamentous bacteria riched in nitrate reached the anoxic zone, where they were exposed to high levels of carbon but limited nitrate, they underwent denitrification. However, when nonfilamentous bacteria were exposed to similar conditions, denitrification was restrained due to their intrinsic nitrate limitation. Hence, in order to avoid filamentous bulking, the nitrate concentration in the return sludge (from aerobic zone to the anoxic zone) should be above 4 mg/L, or alternatively, the nitrate load in the anoxic zone should be kept at levels above 2.7 mg NO(3-)-N/g SS.

  20. Measurement and regulation of the level of a homogeneous plutonium reactor; Mesure et regulation du niveau d'un reacteur homogene au plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Berger, F.; Bertrand, J

    1958-12-01

    Reactivity depends strongly on disturbances of the level of the plutonium solution In the homogeneous reactor. Proserpine has a small cylindrical core, 250 mm diameter, and 10 liters volume. With a view to reducing the dangers due to corrosion and contamination, the solution level in the core is raised by pneumatic pressure. The level is stabilized by means of a regulating system. During critical experiments the variations of the level are less than one hundredth part of a millimeter. (author) [French] Les variations du niveau de la solution de plutonium dans le reacteur homogene Proserpine ont une grosse influence sur la reactivite, car le coeur est petit (10 litres de solution dans un cylindre de diametre 250 mm). En vue de reduire les dangers dus a la corrosion et a la contamination, la commande du volume liquide est pneumatique. Nous avons realise la stabilite du niveau par une regulation qui, dans les essais en regime critique, limite les variations du plan liquide a une fraction de centieme de millimetre. (auteur)

  1. Development of nuclear thermal hydraulic verification tests and evaluation technology - Development of the ultrasonic method for two-phase mixture level measurement in nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Kim, Sang Jae; Kim, Hyung Tae; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    An ultrasonic method is developed for the measurement of the two-phase mixture level in the reactor vessel or steam generator. The ultrasonic method is selected among the several non-nuclear two-phase mixture level measurement methods through two steps of selection procedure. A commercial ultrasonic level measurement method is modified for application into the high temperature, pressure, and other conditions. The calculation method of the ultrasonic velocity is modified to consider the medium as the homogeneous mixture of air and steam, and to be applied into the high temperature and pressure conditions. The cross-correlation technique is adopted as a detection method to reduced the effects of the attenuation and the diffused reflection caused by surface fluctuation. The waveguides are developed to reduce the loss of echo and to remove the effects of obstructs. The present experimental study shows that the developed ultrasonic method measures the two-phase mixture level more accurately than the conventional methods do. 21 refs., 60 figs., 13 tabs. (Author)

  2. Effect of nitrate concentration on filamentous bulking under low level of dissolved oxygen in an airlift inner circular anoxic-aerobic incorporate reactor

    Institute of Scientific and Technical Information of China (English)

    Yiming Su; Yalei Zhang; Xuefei Zhou; Ming Jiang

    2013-01-01

    This laboratory research investigated a possible cause of filamentous bulking under low level of dissolved oxygen conditions (dissolved oxygen value in aerobic zone maintained between 0.6-0.8 mg O2/L) in an airlift inner-circular anoxic-aerobic reactor.During the operating period,it was observed that low nitrate concentrations affected sludge volume index significantly.Unlike the existing hypothesis,the batch tests indicated that filamentous bacteria (mainly Thiothrix sp.) could store nitrate temporarily under carbon restricted conditions.When nitrate concentration was below 4 mg/L,low levels of carbon substrates and dissolved oxygen in the aerobic zone stimulated the nitrate-storing capacity of filaments.When filamentous bacteria riched in nitrate reached the anoxic zone,where they were exposed to high levels of carbon but limited nitrate,they underwent denitrification.However,when nonfilamentous bacteria were exposed to similar conditions,denitrification was restrained due to their intrinsic nitrate limitation.Hence,in order to avoid filamentous bulking,the nitrate concentration in the return sludge (from aerobic zone to the anoxic zone) should be above 4 mg/L,or alternatively,the nitrate load in the anoxic zone should be kept at levels above 2.7 mg NO-3-N/g SS.

  3. Chemical Reactors.

    Science.gov (United States)

    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)

  4. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  5. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  6. Reactor Engineering

    Science.gov (United States)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  7. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  8. Annual report of Power Reactor and Nuclear Fuel Development Corporation, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    The experimental FBR `Joyo` has continued the irradiation operation at 100 MWt. After the 11th periodic inspection, the 30th cycle operation was carried out. The cumulative operation time as of the end of the fiscal year was 51,630 hours, and the cumulative heat output was about 4.2 billion kWh. The prototype FBR `Monju` has succeeded in electric power generation in August, 1995, but the sodium leak accident occurred in December, 1995. The elucidation of the cause of the sodium leak accident and the total inspection for the safety have been carried out. As for FBRs, the research and development of the reactor physics, the design of a large FBR, the equipment systems, the fuel and materials, the structures and the safety have been advanced. The ATR `Fugen` Power Station has continued the operation smoothly, and as of the end of the fiscal year, the total generated electric power was about 17.3 billion kWh, and the capacity factor was 66.3%. It boasts about the result of using MOX fuel. The exploration of uranium resources, the development of uranium conversion, uranium enrichment and plutonium fuel, the reprocessing of spent fuel, the development of environmental technology for radioactive waste, creative and innovative research and development, safety control and safety research and others are reported. (K.I.)

  9. Low-level radioactive waste from commercial nuclear reactors. Volume 4. Proceedings of the workshop on research and development needs for treatment of low-level radioactive waste from commercial nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Godbee, H.W.; Frederick, E.J.; Jolley, R.L.; Kibbey, A.H.; Rodgers, B.R. (comps.)

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. As part of this program, a workshop was conducted for determining research and development needs in LLRW treatment. Volume 4, the proceedings of this workshop, includes the formal presentations and both panel and general discussions dealing with such issues as disposal, compaction, and the ''below regulatory concern'' philosophy. Summaries of individual workshops dealing with specific aspects of LLRW treatment are also presented in this volume.

  10. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool; Determinacion de la rapidez de exposion en el nivel 49.90 del edificio del reactor por la disminucion en el nivel de agua de la alberca de combustible gastado

    Energy Technology Data Exchange (ETDEWEB)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C., E-mail: zoedelfin@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The fuel assemblies storage in the nuclear power plant of Laguna Verde (NPP-L V) represents a crucial aspect, due to the generated dose by the decay heat of the present radio-nuclides in the assemblies retired of the reactor core, after their useful life. These spent assemblies are located inside the spent fuel pool (SFP), in the level 49.90 m in the Reload Floor of the Reactor building of NPP-L V. This leads to the protection at personnel applying the ALARA (As Low As Reasonably Achievable) criteria, fulfilling the established dose criteria by the Regulator Body the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). Considering the loss scenario of the cooling system of the SFP, in which the SFP water vaporizes, is important to know the water level in which the limit of effective dose equivalent is fulfilled for the personnel. Also, is important for the instrumentation of the SFP, for the useful life of the same instruments. In this work is obtained the exposition rapidity corresponding to different water levels of SFP in the Reload Floor of NPP-L V, to identify the minimum level of water where the limit of effective dose equivalent is fulfilled of 25 rem s to the personnel, established in the Article 48 of the General Regulation of Radiological Safety of CNSNS and the Chapter 50 Section 67 of the 10-Cfr of Nuclear Regulatory Commission in USA. The water level is also identified where the exposition rapidity is of 15 m R/hr, being the value of the set point of the area radiation monitor D21-Re-N003-1, located to 125 cm over the level 49.90 meters of the Reload Floor of NPP-L V. (Author)

  11. Hydrodynamics of multi-phase packed bed micro-reactors

    NARCIS (Netherlands)

    Márquez Luzardo, N.M.

    2010-01-01

    Why to use packed bed micro-reactors for catalyst testing? Miniaturized packed bed reactors have a large surface-to-volume ratio at the reactor and particle level that favors the heat- and mass-transfer processes at all scales (intra-particle, inter-phase and inter-particle or reactor level). If the

  12. Hydrodynamics of multi-phase packed bed micro-reactors

    NARCIS (Netherlands)

    Márquez Luzardo, N.M.

    2010-01-01

    Why to use packed bed micro-reactors for catalyst testing? Miniaturized packed bed reactors have a large surface-to-volume ratio at the reactor and particle level that favors the heat- and mass-transfer processes at all scales (intra-particle, inter-phase and inter-particle or reactor level). If the

  13. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  14. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  15. Low-level radioactive waste from commercial nuclear reactors. Volume 2. Treatment, storage, disposal, and transportation technologies and constraints

    Energy Technology Data Exchange (ETDEWEB)

    Jolley, R.L.; Dole, L.R.; Godbee, H.W.; Kibbey, A.H.; Oyen, L.C.; Robinson, S.M.; Rodgers, B.R.; Tucker, R.F. Jr.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 2 discusses the definition, forms, and sources of LLRW; regulatory constraints affecting treatment, storage, transportation, and disposal; current technologies used for treatment, packaging, storage, transportation, and disposal; and the development of a matrix relating treatment technology to the LLRW stream as an aid for choosing methods for treating the waste. Detailed discussions are presented for most LLRW treatment methods, such as aqueous processes (e.g., filtration, ion exchange); dewatering (e.g., evaporation, centrifugation); sorting/segregation; mechanical treatment (e.g., shredding, baling, compaction); thermal processes (e.g., incineration, vitrification); solidification (e.g., cement, asphalt); and biological treatment.

  16. Low-level radioactive waste from commercial nuclear reactors. Volume 1. Recommendations for technology developments with potential to significantly improve low-level radioactive waste management

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, B.R.; Jolley, R.L.

    1986-02-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 1 provides an executive summary and a general introduction to the four-volume set, in addition to recommendations for research and development (R and D) for low-level radioactive waste (LLRW) treatment. Generic, long-range, and/or high-risk programs identified and prioritized as needed R and D in the LLRW field include: (1) systems analysis to develop decision methodology; (2) alternative processes for dismantling, decontaminating, and decommissioning; (3) ion exchange; (4) incinerator technology; (5) disposal technology; (6) demonstration of advanced technologies; (7) technical assistance; (8) below regulatory concern materials; (9) mechanical treatment techniques; (10) monitoring and analysis procedures; (11) radical process improvements; (12) physical, chemical, thermal, and biological processes; (13) fundamental chemistry; (14) interim storage; (15) modeling; and (16) information transfer. The several areas are discussed in detail.

  17. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  18. Photocatalytic degradation of mixed gaseous carbonyl compounds at low level on adsorptive TiO2/SiO2 photocatalyst using a fluidized bed reactor.

    Science.gov (United States)

    Zhang, Maolin; An, Taicheng; Fu, Jiamo; Sheng, Guoying; Wang, Xinming; Hu, Xiaohong; Ding, Xuejun

    2006-06-01

    An adsorptive silica-supported titania photocatalyst TiO(2)/SiO(2) was prepared by using nanosized titania (anatase) immobilized on silica gel by the sol-gel technique with the titanium tetra isopropoxide as the main raw material and acetic acid as the acid catalyst. Meanwhile the structure and properties of the TiO(2)/SiO(2) photocatalyst were studied by means of many modern analysis techniques such as TEM, XRD, and BET. Gas-solid heterogeneous photocatalytic decomposition of four carbonyl compounds mixture at low concentration levels over ultraviolet irradiated TiO(2)/SiO(2) photocatalyst were carried out with high degradation efficiencies in a coaxial triple-cylinder-type fluidized bed photocatalytic reactor, which provided efficient continuous contact of ultraviolet photons, silica-supported titania photocatalyst, and gaseous reactants. Experimental results showed that the photocatalyst had a high adsorption performance and a good photocatalytic activity for four carbonyl compounds mixture. Some factors influencing the photocatalytic decomposition of the mixed carbonyl compounds, i.e. the gas flowrate, relative humidity, concentration of oxygen, and illumination time, were discussed in detail. It is found that the photocatalytic reaction rate of four carbonyl compounds decreased in this order: propionaldehyde, acetone, acetaldehyde and formaldehyde.

  19. Tandem Mirror Reactor Systems Code (Version I)

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  20. Estimative of core damage frequency in IPEN IEA-R1 research reactor due to the initiating events of loss of flow caused by channel blockage and loss of coolant caused by a large rupture in the pipe of the primary circuit - PSA level 1

    Energy Technology Data Exchange (ETDEWEB)

    Hirata, Daniel Massami [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) Sao Paulo, SP (Brazil)

    2011-07-01

    This work applies the methodology of Probabilistic Safety Assessment Level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid caused by large pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, Emergency Core Cooling System (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions in which these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  1. Concept for LEU Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  2. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-12-31

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  3. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-01-01

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  4. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  5. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  6. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  7. Assessment of torsatrons as reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyon, J.F. (Oak Ridge National Lab., TN (United States)); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia))

    1992-12-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

  8. Jules Horowitz Reactor, basic design

    Energy Technology Data Exchange (ETDEWEB)

    Bergamaschi, Y.; Bouilloux, Y.; Chantoin, P.; Guigon, B.; Bravo, X.; Germain, C.; Rommens, M.; Tremodeux, P

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it. It has the ambition to provide the necessary nuclear data and maintain a fission research capacity in Europe after 2010. This capacity should be service-oriented. It will be established in Cadarache. The Jules Horowitz reactor will also: - represent a significant step in term of performances and experimental capabilities, - be designed with a high flexibility, in order to satisfy the current demand from European industry, research and be able to accommodate future requirements, - reach a high level of safety, according to the best current practice. This paper will present the main functionalities and the design options resulting from the 'preliminary design' studies. (authors)

  9. Nuclear reactor neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  10. Reactor and method of operation

    Science.gov (United States)

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  11. The dependence of helium generation rate on nickel content of Fe-Cr-Ni alloys irradiated at high dpa levels in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Oliver, B.M.; Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    With a few exceptions in the literature, it is generally accepted that it is nickel in Fe-Cr-Ni alloys that produces most of the transmutant helium and that the helium generation rate should scale linearly with the nickel content. Surprisingly, this assumption is based only on irradiations of pure nickel and has never been tested in an alloy series. There have also been no extensive tests of the predictions for helium production in alloys in various fast reactors spectra.

  12. Mitigation of corrosion and mass transfer in sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Latge, C. [CEA Cadarache, Dir. de l' Energie Nucleaire, 13 - Saint-Paul-lez-Durance (France); Feron, D. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif-sur-Yvette (France)

    2009-07-01

    Full text of publication follows: Several coolants can be used for the development of the Fast Reactors, as sodium, gas, lead or lead-bismuth eutectic, and have been selected in the Generation IV forum. The high density energy requires a coolant with a very good thermal conductivity. Liquid sodium is such a medium which is liquid between 97.8 up to 880 C at dynamic pressure below 4 bars, and with compatible neutron-physical properties. Its viscosity is comparable to that of water and its compatibility with metallic materials is fairly satisfactory. It is however necessary to keep the conditions of operation within a range such that corrosion is limited. Several materials are suitable for use in liquid sodium reactors, among ferritic and austenitic steels and high temperature alloys with up to 32% nickel contents. The designer has however to consider the mass transfer between materials of different compositions. The exchange and transfer of non-metallic elements such as carbon or nitrogen has to be taken into account. The corrosion mechanisms of austenitic steels have been extensively studied and described in the literature: surface cleaning, austenitic dissolution, formation of a ferrite layer, steady state equilibrium and several models have been proposed: main parameters include oxygen content, sodium velocity and steel temperature. Operating experience has shown that, if there are no cladding failures, the main source of radioactivity in the primary circuit is the activated corrosion products, like {sup 54}Mn, {sup 51}Cr,..., induced by the activation of core materials which are dissolved into the sodium and mainly deposited in the coldest parts of the reactor i.e. the Intermediate Heat Exchanger (IHX) and pumps. Radio-cobalt such as {sup 60}Co are also produced and a low fraction is deposited in primary components. The corrosion rates estimated and the contamination induced by activated corrosion products observed in SFR like Phenix, JOYO, BN600, PFR, EBR2 have

  13. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  14. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  15. NCSU reactor sharing program. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, P.B.

    1997-01-10

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996.

  16. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H.; Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France))

    1992-01-01

    Most of the first generation of fast reactors that were operated at significant power levels employed solid metal fuels. They were constructed in the United States and United Kingdom in the 1950s and included Experimental Breeder Reactor (EBR)-I and -II operated by Argonne National Laboratory, United States, the Enrico Fermi Reactor operated by the Atomic Power Development Associates, United States and DFR operated by the U.K. Atomic Energy Authority (UKAEA). Their paper tracer pre-development of fast reactor fuel from these early days through the 1980s including ceramic fuels.

  17. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  18. Fast reactor irradiation effects on fracture toughness of Si3N4 in comparison with MgAl2O4 and yttria stabilized ZrO2

    Science.gov (United States)

    Tada, K.; Watanabe, M.; Tachi, Y.; Kurishita, H.; Nagata, S.; Shikama, T.

    2016-04-01

    Fracture toughness of silicon nitride (Si3N4), magnesia-alumina spinel (MgAl2O4) and yttria stabilized zirconia (8 mol%Y2O3-ZrO2) was evaluated by the Vickers-indentation technique after the fast reactor irradiation up to 55 dpa (displacement per atom) at about 700 °C in the Joyo. The change of the fracture toughness by the irradiation was correlated with nanostructural evolution by the irradiation, which was examined by transmission electron microscopy. The observed degradation of fracture toughness in Si3N4 is thought to be due to the relatively high density of small-sized of the irradiation induced defects, which should be resulted from a large amount of transmutation gases of hydrogen and helium. Observed improvement of fracture toughness in MgAl2O4 was due to the blocking of crack propagation by the antiphase boundaries. The radiation effects affected the fracture toughness of yttria stabilized zirconia at 55 dpa, suggesting that the generated high density voids would affect the propagation of cracks.

  19. Biotic, temporal and spatial variability of tritium concentrations in transpirate samples collected in the vicinity of a near-surface low-level nuclear waste disposal site and nearby research reactor.

    Science.gov (United States)

    Twining, J R; Hughes, C E; Harrison, J J; Hankin, S; Crawford, J; Johansen, M; Dyer, L

    2011-06-01

    The results of a 21 month sampling program measuring tritium in tree transpirate with respect to local sources are reported. The aim was to assess the potential of tree transpirate to indicate the presence of sub-surface seepage plumes. Transpirate gathered from trees near low-level nuclear waste disposal trenches contained activity concentrations of (3)H that were significantly higher (up to ∼700 Bq L(-1)) than local background levels (0-10 Bq L(-1)). The effects of the waste source declined rapidly with distance to be at background levels within 10s of metres. A research reactor 1.6 km south of the site contributed significant (p nuclear waste site.

  20. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

    losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  1. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  2. Investigation of KW reactor incident

    Energy Technology Data Exchange (ETDEWEB)

    Sturges, D G [USAEC Hanford Operations Office, Richland, WA (United States); Hauff, T W; Greager, O H [General Electric Co., Richland, WA (United States). Hanford Atomic Products Operation

    1955-02-11

    The new KW reactor was placed in operation on January 4, 1955, and had been running at relatively low power levels for only 17 hours when it was shut down because of a process tube water leak which appeared to be associated with a slug rupture. After several days of unrewarding effort to remove the slugs and tube by customary methods, it developed that considerable melting of the tube and slugs had taken place. It was then evident that removal of the stuck mass and repairs to the damaged tube channel would require unusual measures that were certain to extend the reactor outage for several weeks. This report documents the work and findings of the Committee which investigated the KW reactor incident. Its content represents unanimous agreement among the three Committee members.

  3. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  4. Attrition reactor system

    Science.gov (United States)

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  5. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  6. Flow Reactors

    Science.gov (United States)

    The twelve principles of Green Chemistry presented by Anastas and Warner provide the philosophical basis and identify potential areas to increase the level of greenness in designing or implementing chemical reactions in the pharmaceutical industry. With these efforts in mind, the...

  7. Flow Reactors

    Science.gov (United States)

    The twelve principles of Green Chemistry presented by Anastas and Warner provide the philosophical basis and identify potential areas to increase the level of greenness in designing or implementing chemical reactions in the pharmaceutical industry. With these efforts in mind, the...

  8. Development of technology for next generation reactor - Development of next generation reactor in Korea -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    The project, development of next generation reactor, aims overall related technology development and obtainment of related license in 2001. The development direction is to determine the reactor type and to build up the design concept in 1994. For development trend analysis of foreign next generation reactor, level-1 PSA, fuel cycle analysis and computer code development are performed on System 80+ and AP 600. Especially for design characteristics analysis and volume upgrade of AP 600, nuclear fuel and reactor core design analysis, coolant circuit design analysis, mechanical structure design analysis and safety analysis etc. are performed. (Author).

  9. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  10. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  11. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS); Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  12. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  13. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  14. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  15. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  16. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-15

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  17. Simulation of a turbine trip from maximum power level without reactor trip in the TRILLO plant with the code TRACE v5.0 p3; Simulacion de un disparo de turbina desde maximo nivel de potencia sin disparo del reactor en la planta de TRILLO con el codigo TRACE v5.0 p3

    Energy Technology Data Exchange (ETDEWEB)

    Berna, C.; Escriva, A.; Munoz-Cobo, J. L.; Posada, J. M.

    2014-07-01

    The work consists in the simulation of code TRACE v5.0 p3 of the transient in turbine trip from highest level of power without reactor trip. In particular, a steady state with conditions very similar to the of the previous simulation made using the RELAP-MOD3 code has been obtained. In the transient, has been also satisfactory results, specifically the values of pressures, temperatures and mass flows, both in the secondary and primary circuit flow, are also very similar in both cases. In conclusion, have shown the ability to play the transition in study by the TRILLO plant using the code TRACE v5.0 p3 model, constituting a step in the process of verification of such a code. (Author)

  18. NCSU Reactor Sharing Program. Final technical report, [September 1, 1980--August 29, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Perez, P.B.

    1993-11-10

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities.

  19. The role of inoculum and reactor configuration for microbial community composition and dynamics in mainstream partial nitritation anammox reactors.

    Science.gov (United States)

    Agrawal, Shelesh; Karst, Søren M; Gilbert, Eva M; Horn, Harald; Nielsen, Per H; Lackner, Susanne

    2017-03-10

    Implementation of partial nitritation anammox (PNA) in the mainstream (municipal wastewater treatment) is still under investigation. Microbial community structure and reactor type can influence the performance of PNA reactor; yet, little is known about the role of the community composition of the inoculum and the reactor configuration under mainstream conditions. Therefore, this study investigated the community structure of inocula of different origin and their consecutive community dynamics in four different lab-scale PNA reactors with 16S rRNA gene amplicon sequencing. These reactors were operated for almost 1 year and subjected to realistic seasonal temperature fluctuations as in moderate climate regions, that is, from 20°C in summer to 10°C in winter. The sequencing analysis revealed that the bacterial community in the reactors comprised: (1) a nitrifying community (consisting of anaerobic ammonium-oxidizing bacteria (AnAOB), ammonia-oxidizing bacteria (AOB), and nitrite-oxidizing bacteria (NOB)); (2) different heterotrophic denitrifying bacteria and other putative heterotrophic bacteria (HB). The nitrifying community was the same in all four reactors at the genus level, although the biomasses were of different origin. Community dynamics revealed a stable community in the moving bed biofilm reactors (MBBR) in contrast to the sequencing batch reactors (SBR) at the genus level. Moreover, the reactor design seemed to influence the community dynamics, and reactor operation significantly influenced the overall community composition. The MBBR seems to be the reactor type of choice for mainstream wastewater treatment.

  20. Criticality in a high level waste repository. A review of some important factors and an assessment of the lessons that can be learned from the Oklo reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M. [VMO Konsult, Stockholm (Sweden)

    1996-06-01

    The conditions and scenarios that might allow sufficient {sup 239}Pu and/or {sup 235}U to accumulate together with enough water to allow for moderation of neutron energies and thereby achieving a state where neutron-induced fission reactions could be sustained at a rate significantly above the natural rate of spontaneous fission is discussed. The uranium deposit in Oklo, Gabon, which was the site of naturally-occurring neutron-induced fission reactions approximately 2000 My ago is described. The chemistry, mineralogy, and conditions of the nuclear reactor operations are reviewed. Results of modelling the conditions for criticality at Oklo are used to estimate the amounts of spent fuel uranium that must be assembled in a favorable geometry in order to produce a similar reactive situation in a geologic repository. The amounts of uranium that must be transported and redeposited to reach a critical configuration are extremely large in relation to those that could be transported under any reasonably achievable conditions. In addition, transport and redeposition scenarios often require opposite chemical characteristics. It is concluded that the likelihood of achieving a critical condition due to accumulation of a critical mass of uranium outside the canisters after disposal is negligible. Criticality inside the canister is rendered impossible by the use of low-solubility materials inside the canisters that fill space and prevent the entry of enough water to allow moderation of neutron energies. Criticality due to plutonium outside the canister can be ruled out because it requires a series of processes, each of which has a vanishingly small probability. 25 refs, 9 tabs, 8 figs.

  1. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  2. Greater-than-Class C low-level waste characterization. Appendix G: Evaluation of potential for greater-than-Class C classification of irradiated hardware generated by utility-operated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cline, J.E.

    1991-08-01

    This study compiles and evaluates data from many sources to expand a base of data from which to estimate the activity concentrations and volumes of greater-than-Class C low-level waste that the Department of Energy will receive from the commercial power industry. Sources of these data include measurements of irradiated hardware made by or for the utilities that was classified for disposal in commercial burial sites, measurements of neutron flux in the appropriate regions of the reactor pressure vessel, analyses of elemental constituents of the particular structural material used for the components, and the activation analysis calculations done for hardware. Evaluations include results and assumptions in the activation analyses. Sections of this report and the appendices present interpretation of data and the classification definitions and requirements.

  3. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  4. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  5. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  6. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  7. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  8. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  9. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  10. System assessment of helical reactors in comparison with tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-{beta}{sub N} tokamak reactors. (author)

  11. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  12. Nonlinear adaptive power-level control for modular high temperature gas-cooled reactors%模块式高温气冷堆的非线性自适应功率控制

    Institute of Scientific and Technical Information of China (English)

    董哲

    2013-01-01

    Because of its inherent safety feature and potential economic competitiveness,the modular high temperature gas-cooled reactor (MHTGR) is regarded as the central part of the next generation of nuclear plant(NGNP).Power-level control is one of the key techniques that guarantee the safe,stable and efficient operation for any nuclear reactors.Since the MHTGR dynamics have the characteristics of strong nonlinearity and uncertainty,which can improve the operation performance.It is significant to develop the nonlinear adaptive power-level control er for the MHTGR.Based on the natural dynamic features beneficial to system stabilization,a novel nonlinear adaptive power-level controller is given for the MHTGR in this paper.It has proved theoretical y that this newly-built control er not only provides the overal asymptotic closed-loop stability,but also is adaptive to the system uncertainty.This control er has been applied in the power-level regulation of the pebble-bed MHTGR of the HTR-PM power plant.Numerical simulation results confirm the feasibility of this control strategy and the relationship between the performance and parameters.%由于具有固有安全性和潜在经济竞争力,模块式高温气冷堆(Modular High Temperature Gas-cooled Reactor, MHTGR)已被视为下一代核能系统的首选堆型之一。功率调节是保证MHTGR安全、稳定和高效运行的关键技术之一,而MHTGR在动态特性上具有非线性强和不确定性大的特点,因此发展MHTGR的非线性自适应功率控制器对于提升运行性能具有重要意义。在充分利用系统自身对镇定有益的动态特性的基础上,本文给出一种新颖的MHTGR非线性自适应功率控制器,并从理论上证明该控制器既可以保证闭环系统的全局渐近稳定性,又对系统不确定性具有自适应功能。数值仿真结果不仅验证了控制器的可行性,而且还给出了调节性能与控制器参数间的关系。

  13. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  14. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    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.

  15. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  16. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  17. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  18. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  19. On-line solid-phase enrichment coupled to packed reactor flow injection analysis in a green analytical procedure to determine low levels of folic acid using fluorescence detection

    Directory of Open Access Journals (Sweden)

    Emara Samy

    2012-12-01

    Full Text Available Abstract Background Analysis of folic acid (FA is not an easy task because of its presence in lower concentrations, its lower stability under acidic conditions, and its sensitiveness against light and high temperature. The present study is concerned with the development and validation of an automated environmentally friendly pre-column derivatization combined by solid-phase enrichment (SPEn to determine low levels of FA. Results Cerium (IV trihydroxyhydroperoxide (CTH as a packed oxidant reactor has been used for oxidative cleavage of FA into highly fluorescent product, 2-amino-4-hydroxypteridine-6-carboxylic acid. FA was injected into a carrier stream of 0.04 M phosphate buffer, pH 3.4 at a flow-rate of 0.25 mL/min. The sample zone containing the analyte was passed through the CTH reactor thermostated at 40°C, and the fluorescent product was trapped and enriched on a head of small ODS column (10 mm x 4.6 mm i.d., 5 μm particle size. The enriched product was then back-flush eluted by column-switching from the small ODS column to the detector with a greener mobile phase consisting of ethanol and phosphate buffer (0.04M, pH 3.4 in the ratio of 5:95 (v/v. The eluent was monitored fluorimetrically at emission and excitation wavelengths of 463 and 367 nm, respectively. The calibration graph was linear over concentrations of FA in the range of 1.25-50 ng/mL, with a detection limit of 0.49 ng/mL. Conclusion A new simple and sensitive green analytical procedure including on-line pre-column derivatization combined by SPEn has been developed for the routine quality control and dosage form assay of FA at very low concentration level. The method was a powerful analytical technique that had excellent sensitivity, sufficient accuracy and required relatively simple and inexpensive instrumentation.

  20. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  1. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  2. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  3. Nanoscale Polysulfides Reactors Achieved by Chemical Au-S Interaction: Improving the Performance of Li-S Batteries on the Electrode Level.

    Science.gov (United States)

    Fan, Chao-Ying; Xiao, Pin; Li, Huan-Huan; Wang, Hai-Feng; Zhang, Lin-Lin; Sun, Hai-Zhu; Wu, Xing-Long; Xie, Hai-Ming; Zhang, Jing-Ping

    2015-12-23

    In this work, the chemical interaction of cathode and lithium polysulfides (LiPSs), which is a more targeted approach for completely preventing the shuttle of LiPSs in lithium-sulfur (Li-S) batteries, has been established on the electrode level. Through simply posttreating the ordinary sulfur cathode in atmospheric environment just for several minutes, the Au nanoparticles (Au NPs) were well-decorated on/in the surface and pores of the electrode composed of commercial acetylene black (CB) and sulfur powder. The Au NPs can covalently stabilize the sulfur/LiPSs, which is advantageous for restricting the shuttle effect. Moreover, the LiPSs reservoirs of Au NPs with high conductivity can significantly control the deposition of the trapped LiPSs, contributing to the uniform distribution of sulfur species upon charging/discharging. The slight modification of the cathode with <3 wt % Au NPs has favorably prospered the cycle capacity and stability of Li-S batteries. Moreover, this cathode exhibited an excellent anti-self-discharge ability. The slight decoration for the ordinary electrode, which can be easily accessed in the industrial process, provides a facile strategy for improving the performance of commercial carbon-based Li-S batteries toward practical application.

  4. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  5. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  6. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  7. Process Design and Commissioning Operation of Constant Water-Level Sequencing Batch Reactor (CWSBR) at low Temperature%低温环境下CWSBR工艺设计与调试运行

    Institute of Scientific and Technical Information of China (English)

    马海龙; 沙瑛; 陈永

    2012-01-01

    The process of constant water-level sequencing batch reactor (CWSBR) was used in Dongwuqi wastewater treatment plant in Inner Mongolia province. After commissioning operation, nutrient removal effect was realized in the system at low temperature (8 - 10 t). The result* showed that the effluent qualities were qualified to Standard I-A of "Discharge Standard of Pollutants for Municipal Wastewater Treatment Plant" (CB 18918-2002) with CWSBR at low temperature.%内蒙古东乌珠旗污水处理厂采用恒水位SBR工艺处理城市污水,冬季运行的温度为8~10℃,通过调试运行,实现了系统低温条件下脱氮除磷的效果.结果表明:CWSBR工艺在低温条件下保证了出水水质符合GB18918-2002《城镇污水处理厂污染物排放标准》的一级A标准.

  8. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  9. 高功率MPCVD金刚石膜透波窗口材料制备研究%Synthesis of Diamond Film as Electromagnetic Window Material by an Ellipsoidal MPCVD Reactor at High Input Microwave Power Levels

    Institute of Scientific and Technical Information of China (English)

    于盛旺; 刘艳青; 唐伟忠; 申艳艳; 贺志勇; 唐宾

    2012-01-01

    使用自行研制的椭球谐振腔式MPCVD装置,以H2-CH4为气源,在沉积功率8 kW条件下,对大面积金刚石膜透波窗口材料进行了制备研究.分别使用扫描电镜、Raman、分光光谱仪、热导率测试仪和空腔谐振法对金刚石膜的表面形貌、品质、光透过率、热导率和微波复介电常数等进行了表征及测试.实验结果表明,使用自行研制的椭球谐振腔式MPCVD装置,能够满足较高功率下高品质金刚石膜的快速沉积;抛光后的自支撑金刚石膜具有高的光学透过率和热导率,在23 ~ 36 GHz频率范围内微波介电损耗小于1×10-4,有着良好的微波介电性能,是较为理想的透波窗口材料.%Polycrystalline diamond film was grown using H2-CH4 as the source gas in a newly developed ellipsoidal MPCVD reactor at 8 kW microwave power levels. Surface morphology, quality, transmittance, thermal conductivity as well as complex permittivity of the diamond film were examined by scanning electron microscopy, Raman spectroscopy, spectrophotometer, thermal constant tester and cavity resonator method respectively. The result proved that the new ellipsoidal microwave plasma CVD reactor has the ability to deposit high quality diamond films at a high synthesis rate. After polished on both sides, the free-standing diamond film has high transmittance and thermal conductivity. At the same time the film has excellent microwave dielectric properties for complex permittivity less than 1 x 10-4 in the range of 23-36 GHz, this show the film is one of the perfect electromagnetic window materials.

  10. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iwashige, Kengo

    1996-06-21

    In an LMFBR type reactor, partitions are disposed to a coolant channel at positions lower than the free liquid level, and the width of the partitions is adapted to have a predetermined condition. Namely, when low temperature fluid overflowing the wall of the coolant channel, flows down and collided against the free liquid surface in the coolant channel, since the dropping speed thereof is reduced abruptly, large pressure waves are caused by kinetic force of the low temperature fluid. However, if appropriate numbers of partitions having an appropriate shape are formed, the dropping speed of the low temperature fluid is moderated to reduce the pressure waves. In addition, since the pressure waves are dispersed to the circumferential and lateral directions of the coolant flow channel respectively, the propagation of the pressure waves can be prevented effectively. Further, when the flow of the low temperature fluid is changed to the circumferential direction, for example, by earthquakes, since the partitions act as members resisting against the circumferential change of the low temperature fluid, the change of the direction can be suppressed. (N.H.)

  11. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors, for example, such characteristics include rapid on-line refueling, and a core design with room for such a large number of assemblies or targets that it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors, such as hot cells, where plutonium could be separated, could pose a safeguards challenge because, in some cases, they are not declared (because they are not located in the facility or because nuclear materials are not foreseen to be processed inside) and may not be accessible to inspectors in States without an Additional Protocol in force.

  12. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  13. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  14. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  15. Future Reactor Experiments

    OpenAIRE

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measure...

  16. Reactor Neutrino Experiments

    OpenAIRE

    Cao, Jun

    2007-01-01

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measu...

  17. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  18. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  19. INVAP's Research Reactor Designs

    OpenAIRE

    Eduardo Villarino; Alicia Doval

    2011-01-01

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  20. The reactor antineutrino anomalies

    Energy Technology Data Exchange (ETDEWEB)

    Haser, Julia; Buck, Christian; Lindner, Manfred [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

    2016-07-01

    Major discoveries were made in the past few years in the field of neutrino flavour oscillation. Nuclear reactors produce a clean and intense flux of electron antineutrinos and are thus an essential neutrino source for the determination of oscillation parameters. Most currently the reactor antineutrino experiments Double Chooz, Daya Bay and RENO have accomplished to measure θ{sub 13}, the smallest of the three-flavour mixing angles. In the course of these experiments two anomalies emerged: (1) the reanalysis of the reactor predictions revealed a deficit in experimentally observed antineutrino flux, known as the ''reactor antineutrino anomaly''. (2) The high precision of the latest generation of neutrino experiments resolved a spectral shape distortion relative to the expected energy spectra. Both puzzles are yet to be solved and triggered new experimental as well as theoretical studies, with the search for light sterile neutrinos as most popular explanation for the flux anomaly. This talk outlines the two reactor antineutrino anomalies. Discussing possible explanations for their occurrence, recent and upcoming efforts to solve the reactor puzzles are highlighted.

  1. Moon base reactor system

    Science.gov (United States)

    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.

  2. Reactor and process design in sustainable energy technology

    CERN Document Server

    Shi, Fan

    2014-01-01

    Reactor Process Design in Sustainable Energy Technology compiles and explains current developments in reactor and process design in sustainable energy technologies, including optimization and scale-up methodologies and numerical methods. Sustainable energy technologies that require more efficient means of converting and utilizing energy can help provide for burgeoning global energy demand while reducing anthropogenic carbon dioxide emissions associated with energy production. The book, contributed by an international team of academic and industry experts in the field, brings numerous reactor design cases to readers based on their valuable experience from lab R&D scale to industry levels. It is the first to emphasize reactor engineering in sustainable energy technology discussing design. It provides comprehensive tools and information to help engineers and energy professionals learn, design, and specify chemical reactors and processes confidently. Emphasis on reactor engineering in sustainable energy techn...

  3. RSMASS-D models: An improved method for estimating reactor and shield mass for space reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1997-10-01

    Three relatively simple mathematical models have been developed to estimate minimum reactor and radiation shield masses for liquid-metal-cooled reactors (LMRs), in-core thermionic fuel element (TFE) reactors, and out-of-core thermionic reactors (OTRs). The approach was based on much of the methodology developed for the Reactor/Shield Mass (RSMASS) model. Like the original RSMASS models, the new RSMASS-derivative (RSMASS-D) models use a combination of simple equations derived from reactor physics and other fundamental considerations, along with tabulations of data from more detailed neutron and gamma transport theory computations. All three models vary basic design parameters within a range specified by the user to achieve a parameter choice that yields a minimum mass for the power level and operational time of interest. The impact of critical mass, fuel damage, and thermal limitations are accounted for to determine the required fuel mass. The effect of thermionic limitations are also taken into account for the thermionic reactor models. All major reactor component masses are estimated, as well as instrumentation and control (I&C), boom, and safety system masses. A new shield model was developed and incorporated into all three reactor concept models. The new shield model is more accurate and simpler to use than the approach used in the original RSMASS model. The estimated reactor and shield masses agree with the mass predictions from separate detailed calculations within 15 percent for all three models.

  4. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  5. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  6. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  7. Kinetics for a membrane reactor reducing perchlorate.

    Science.gov (United States)

    Padhye, Lokesh; Rainwater, Ken; Jackson, W Andrew; Morse, Audra

    2007-02-01

    The major objectives of this work were to operate and construct an autohydrogenotrophic reactor and estimate perchlorate degradation kinetics. The results show that autohydrogenotrophic bacteria were cultured in the reactor and capable of removing 3.6 mg/d of perchlorate in the presence of excess hydrogen (99% removal). The reactor was successful in treating the average influent perchlorate concentration of 532 microg/L to the level of 3 microg/L. A first-order relationship was obtained between the concentration of active biomass in the reactor and the hydraulic retention time for the given amount of substrate. During the kinetic loading study, perchlorate removal ranged from 100 to 50%. The kinetic rate of perchlorate degradation observed in this study was 1.62 hr(-1). The significant degradation of perchlorate in these samples indicates the ubiquity of perchlorate-reducing organisms. Additionally, nitrate was simultaneously removed during water treatment (greater than 90% removal). Because of the excess levels of hydrogen, simultaneous removal of nitrate was not believed to significantly affect perchlorate removal. The area of concern was the lack of complete control over biological treatment. The growth of sulfate-reducing organisms in the reactor negatively affected perchlorate removal efficiency. There were no significant effects observed on the dissolved organic carbon and total suspended solids concentration of the effluent, suggesting that the treatment did not produce a large amount of biomass washout.

  8. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

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

    Science.gov (United States)

    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...

  10. Molten salt reactor: Deterministic safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, Elsa; Heuer, Daniel; Mathieu, Ludovic; Le Brun, Christian [Laboratory for Subatomic Physics and Cosmology (LPSC), 53, Avenue des Marthyrs, F-38026 Grenoble (France)

    2006-07-01

    Molten Salt Reactors (MSRs) are one of the systems retained by Generation IV as a candidate for the next generation of nuclear reactors. This type of reactor is particularly well adapted to the thorium fuel cycle (Th- {sup 233}U) which has the advantage of producing less minor actinides than the uranium-plutonium fuel cycle ({sup 238}U- {sup 239}Pu). In the frame of a major re-evaluation of the MSR concept and concentrating on some major constraints such as feasibility, breeding capability and, above all, safety, we have considered a particular reactor configuration that we call the 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum. This reactor is presented in the first section. MSRs benefit from several specific advantages which are listed in a second part of this work. Beyond these advantages of the MSR, the level of the deterministic safety in such a reactor has to be assessed precisely. In a third section, we first draw up a list of the reactivity margins in our reactor configuration. We then define and quantify the parameters characterizing the deterministic safety of any reactor: the fraction of delayed neutrons, and the system's feedback coefficients that are here negative. Finally, using a simple point-kinetic evaluation, we analyze how these safety parameters impact the system when the total reactivity margins are introduced in the MSR. The results of this last study are discussed, emphasizing the satisfactory behavior of the MSR and the excellent level of deterministic safety which can be achieved. This work is based on the coupling of a neutron transport code called MCNP with a materials evolution code. The former calculates the neutron flux and the reaction rates in all the cells while the latter solves the Bateman equations for the evolution of the materials composition within the cells. These calculations take into account the input parameters (power released

  11. The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

    Science.gov (United States)

    Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

    2009-09-01

    In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

  12. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  13. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  14. The RES Reactor. A test reactor for the French naval propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, Sylvestre [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Minguet, Jean-Luc [AREVA-Technicatome, BP17, 91192 Gif-sur-Yvette (France)

    2006-07-01

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  15. Operation of Reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    3.1 Annual Report of SPR Operation Chu Shaochu Having overseen by National Nuclear Safety Administration and specialists, the reactor restarted up successfully after Safety renovation on April 16, 1996. In August 1996 the normal operation of SPR was approved by the authorities of Naitonal Nuclear Safety Administration. 1 Operation status In 1996, the reactor operated safely for 40 d and the energy released was about 137.3 MW·d. The operation status of SPR is shown in table 1. The reactor started up to higher power (power more than 1 MW) and lower power (for physics experiments) 4 times and 14 times respectively. Measurement of control rod efficiency and other measurement tasks were 2 times and 5 times respectively.

  16. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  17. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  18. NGNP Reactor Coolant Chemistry Control Study

    Energy Technology Data Exchange (ETDEWEB)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  19. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  20. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  1. Stade NPP. Dismantling of the reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Scharf, Daniel; Dziwis, Joachim [E.ON Anlagenservice GmbH Nukleartechnik, Gelsenkirchen (Germany); Kemp, Lutz-Hagen [KKW Stade GmbH und Co. oHG, Stade (Germany)

    2012-11-01

    Within the scope of the 4{sup th} partial decommissioning permission of Stade NPP the activated and contaminated structures of the reactor pool had to be dismantled in order to gain a completely non-radioactive reactor pool area for the subsequent clearance measurement of the reactor building. In order to achieve the aim it was intended to remove the activated pool liner sheets, its activated framework and several contaminated ventilation channels made of stainless steel, the concrete walls of the reactor pool entirely or in parts depending on their activation level, as well as the remaining activated carbon steel structures of the reactor pool bottom. Embedded in the concrete walls there were several highly contaminated excore tubes and the contaminated pool top edge, which were intended to be removed to its full extent. The contract of the Stade NPP initiated reactor pool dismantling project had been awarded to E.ON Anlagenservice GmbH (EAS) and its subsupplier sat. Kerntechnik GmbH for the concrete dismantling works and was performed as follows. In order to minimize the radiation level in the main working area in accordance with the ALARA principle, the liner sheets and middle parts of its framework were removed by means of angle grinders first, as they were the most dose rate relevant parts. As a result the primary average radiation level in the reactor pool (measured in a distance of 500 mm from the walls) was lowered from 40 {mu}Sv/h to less than 2 {mu}Sv/h. After the minimization of the radiation level in the working area the main dismantling step started with the cutting of the reactor pool walls in blocks by means of diamond rope cutters. Once a concrete block was cut out, it was transported into the fuel pool by means of a crane and crane fork, examined radiologically, marked area by area and segmented to debris by means of an electrical excavator with a hydraulic chisel. Afterwards the debris and carbon steel parts were fractioned and packed for further

  2. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  3. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  4. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

  5. Low-level radioactive waste from commercial nuclear reactors. Volume 3. Bibliographic abstracts of significant source documents. Part 1. Open-literature abstracts for low-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Bowers, M.K.; Rodgers, B.R.; Jolley, R.L.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 3 of this series is a collection of abstracts of most of the reference documents used for this study. Because of the large volume of literature, the abstracts have been printed in two separate parts. Volume 3, part 1 presents abstracts of the open literature relating to LLRW treatment methodologies. Some of these references pertain to treatment processes for hazardous wastes that may also be applicable to LLRW management. All abstracts have been limited to 21 lines (for brevity), but each abstract contains sufficient information to enable the reader to determine the potential usefulness of the source document and to locate each article. The abstracts are arranged alphabetically by author or organization, and indexed by keyword.

  6. Low-level radioactive waste from commercial nuclear reactors. Volume 3. Bibliographic abstracts of significant source documents. Part 1. Open-literature abstracts for low-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Bowers, M.K.; Rodgers, B.R.; Jolley, R.L.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 3 of this series is a collection of abstracts of most of the reference documents used for this study. Because of the large volume of literature, the abstracts have been printed in two separate parts. Volume 3, part 1 presents abstracts of the open literature relating to LLRW treatment methodologies. Some of these references pertain to treatment processes for hazardous wastes that may also be applicable to LLRW management. All abstracts have been limited to 21 lines (for brevity), but each abstract contains sufficient information to enable the reader to determine the potential usefulness of the source document and to locate each article. The abstracts are arranged alphabetically by author or organization, and indexed by keyword.

  7. Thermionic reactors for space nuclear power

    Science.gov (United States)

    Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

    Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

  8. Risk Management for Sodium Fast Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-01-01

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

  9. Gaseous fuel reactor systems for aerospace applications

    Science.gov (United States)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  10. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  11. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

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

  12. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  13. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  14. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Worrall, Andrew [ORNL; Todosow, Michael [Brookhaven National Laboratory (BNL)

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

  15. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  16. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  17. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  18. MULTISTAGE FLUIDIZED BED REACTOR

    Science.gov (United States)

    Jonke, A.A.; Graae, J.E.A.; Levitz, N.M.

    1959-11-01

    A multistage fluidized bed reactor is described in which each of a number of stages is arranged with respect to an associated baffle so that a fluidizing gas flows upward and a granular solid downward through the stages and baffles, whereas the granular solid stopsflowing downward when the flow of fluidizing gas is shut off.

  19. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  20. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  1. The First Reactor.

    Science.gov (United States)

    Department of Energy, Washington, DC.

    On December 2, 1942, in a racquet court underneath the West Stands of Stagg Field at the University of Chicago, a team of scientists led by Enrico Fermi created the first controlled, self-sustaining nuclear chain reaction. This updated and revised story of the first reactor (or "pile") is based on postwar interviews (as told to Corbin…

  2. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  3. Chromatographic and Related Reactors.

    Science.gov (United States)

    1988-01-07

    special information about effects of surface heteroge- neity in the methanation reaction. Studies of an efficient multicolumn assembly for measuring...of organic basic catalysts such as pyridine and 4-methylpicoline. It was demonstrated that the chromatographic reactor gave special information about...Programmed Reaction to obtain special information about surface heterogeneity in the methanation reaction. Advantages of stopped flow over steady state

  4. New concepts for shaftless recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.; Berty, I.J.

    1987-01-01

    Berty Reaction Engineers, Ltd. (BREL) is developing two new laboratory recycle reactors, the ROTOBERTY and the TURBOBERTY. These new reactors are basically improved versions of the original Berty reactor. To understand why the reactors have the features that they do, it is first necessary to briefly review laboratory reactors in general and specifically the original Berty reactor.

  5. Brazilian multipurpose reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

  6. Modeling Chemical Reactors I: Quiescent Reactors

    CERN Document Server

    Michoski, C E; Schmitz, P G

    2010-01-01

    We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

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

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

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

  8. Study on the License Requirements for the SRO/RO of the Kijang Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Subeom; Shin, Taemyung [Korea Nat. University of Transportation, Seoul (Korea, Republic of); Chae, H. T.; Ahn, G. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, S. J.; Gam, S. C. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of the study is to propose an appropriate regulatory position for the Kijang reactor operator license requirement by the review of the applicability and compatibility of HANARO SRO/RO license holders for Kijang reactor operation. As the area using radioactive isotope became gradually enlarged both inside and outside of the country, the Kijang research reactor is planned and now under construction next to the HANARO research reactor now being operated in Taejon. In this paper, therefore, an establishment of revised operator license system is discussed for the new research reactor. The design and operation characteristics of the two (HANARO and Kijang) reactors are concluded to be very similar to each other, however, there still exist slight differences in some minor portions. It is recommendable to allow an independent license for each reactor if two reactors of the same power level have recognizable differences in the design and operation characteristics.

  9. Reactor monitoring using antineutrino detectors

    Science.gov (United States)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  10. Reactor vessel support system. [LMFBR

    Science.gov (United States)

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  12. Target-fueled nuclear reactor for medical isotope production

    Science.gov (United States)

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

  13. High-temperature reactor developments in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Schram, R.P.C.; Cordfunke, E.H.P.; Heek, A.I. van

    1996-01-01

    The high-temperature reactor development in the Netherland is embedded in the WHITE reactor program, in which several Dutch research institutes and engineering companies participate. The activities within the WHITE program are focused on the development of a small scale HTS for combined heat and power generation. In 1995, design choices for a pebble bed reactor were made at ECN. The first concept HTR will gave a closed cycle helium turbine and a power level of 40 MWth. It is intended to make the market introduction of a commercially competitive HTR feasible. The design will be an optimization of the Peu-a-Peu (PAP) concept of KFA Juelich. Computer codes necessary for the evaluation of reactor physics aspects of this reactor are developed in cooperation with international partners. An evaluation of a 20 MWth PAP concept showed that the maximum fuel termmperature after depressurization does not exceed 1300 C. (orig.).

  14. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  15. Antineutrino monitoring for the Iranian heavy water reactor

    CERN Document Server

    Christensen, Eric; Jaffke, Patrick; Shea, Thomas

    2014-01-01

    In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

  16. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  17. Decommissioning of the high flux beam reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J.P. [National Synchrotron Light Source, Brookhaven Laboratory, Upton, NY 11973 (United States); Reciniello, R.N. [Radiological Control Div., Brookhaven Laboratory, Upton, NY 11973 (United States); Holden, N.E. [National Nuclear Data Center, Brookhaven Laboratory, Upton, NY 11973 (United States)

    2011-07-01

    The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Dept. of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996-2009 is described in the paper. (authors)

  18. Detection of anomalous reactor activity using antineutrino count evolution over the course of a reactor cycle

    Science.gov (United States)

    Bulaevskaya, Vera; Bernstein, Adam

    2011-06-01

    This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard "baseline" fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 82 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be obtained by various means, including use of the method in conjunction with existing reactor safeguards methods. We also identify a necessary and sufficient minimum daily antineutrino count rate and a maximum tolerable background rate to achieve the quoted sensitivity, and list examples of detectors in which such rates have been attained.

  19. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  20. MEANS FOR COOLING REACTORS

    Science.gov (United States)

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  1. Integrated Microfluidic Reactors.

    Science.gov (United States)

    Lin, Wei-Yu; Wang, Yanju; Wang, Shutao; Tseng, Hsian-Rong

    2009-12-01

    Microfluidic reactors exhibit intrinsic advantages of reduced chemical consumption, safety, high surface-area-to-volume ratios, and improved control over mass and heat transfer superior to the macroscopic reaction setting. In contract to a continuous-flow microfluidic system composed of only a microchannel network, an integrated microfluidic system represents a scalable integration of a microchannel network with functional microfluidic modules, thus enabling the execution and automation of complicated chemical reactions in a single device. In this review, we summarize recent progresses on the development of integrated microfluidics-based chemical reactors for (i) parallel screening of in situ click chemistry libraries, (ii) multistep synthesis of radiolabeled imaging probes for positron emission tomography (PET), (iii) sequential preparation of individually addressable conducting polymer nanowire (CPNW), and (iv) solid-phase synthesis of DNA oligonucleotides. These proof-of-principle demonstrations validate the feasibility and set a solid foundation for exploring a broad application of the integrated microfluidic system.

  2. Reactor Neutrino Spectra

    OpenAIRE

    Hayes, A. C.; Vogel, Petr

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these spectra and their associated uncertainties is crucial for neutrino oscillation studies. The spectra used to date have been determined either by converting measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that make up the spectra, using modern databases as input. The uncertainties in the subdominant corrections to β-decay plague both methods, and we ...

  3. REACTOR MODERATOR STRUCTURE

    Science.gov (United States)

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  4. BOILER-SUPERHEATED REACTOR

    Science.gov (United States)

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  5. The OPAL reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.; Irwin, T. [Australian Nuclear Science and Technology Organisation, Sydney (Australia); Ordonez, J.P. [INVAP SE, Bariloche (Argentina)

    2007-07-01

    The project to provide a replacement for Australia's HIFAR reactor began with governmental approval in September 1997 and reached its latest milestone with the achievement of the first full power operation of the OPAL reactor in November 2006. OPAL is a pool-type reactor with a thermal power of 20 MW and a fuel enrichment maximum of 20 per cent. This has been a successful project for both ANSTO (Australian Nuclear Science and Technology Organisation) and the contractor INVAP SE. This project was characterised by extensive interaction with the project's stake-holders during project definition and the use of a performance-based turnkey contract which gave the contractor the maximum opportunity to optimise the design to achieve performance and cost effectiveness. The contactor provided significant in-house resources as well as capacity to manage an international team of suppliers and sub-contractors. A key contributor to the project's successful outcomes has been the development and maintenance of an excellent working relationship between ANSTO and INVAP project teams. Commissioning was undertaken in accordance with the IAEA recommended stages. This paper presents the approaches used to define the project requirements, to choose the supplier and to deliver the project. The main results of hot commissioning are reviewed and the problems encountered examined. Operational experience since hot commissioning is also reviewed.

  6. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  7. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  8. Optimized transition from the reactors of second and third generations to the thorium molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, E.; Heuer, D.; Allibert, M.; Ghetta, V.; Le Brun, C.; Mathieu, L.; Brissot, R.; Liatard, E. [Laboratoire de Physique Subatomique et de Cosmologie (LPSC/IN2P3/CNRS), 38 - Grenoble (France)

    2007-07-01

    We present in this article a very promising, simple and feasible concept of Molten Salt Reactor with no moderator in the core, operated in the Th/U{sup 233} fuel cycle with fluoride salts and called non-moderated Thorium Molten Salt Reactor (TMSR). We have detailed in this article some parametric studies, related to the system reprocessing constraints, and the heavy nuclei composition of the salt which modifies the neutron spectrum of the reactor. Since U{sup 233} does not exist on earth and is not being produced today, we aim at designing a critical MSR able to burn the Plutonium and the Minor Actinides produced in the current operating reactors, and consequently to convert this Plutonium into U{sup 233}. This leads to closing the current fuel cycle thanks to TMSRs started with transuranic elements on a Thorium base, i.e. started in the Th/Pu fuel cycle, similarly to fast neutron reactors operated in the U/Pu fuel cycle. The burning of transuranic elements in these Pu-started TMSRs results in high waste reduction rates, up to 95-97% for all TMSR configurations assessed. We particularly point out in our analyses the excellent level of deterministic safety of all the TMSR configurations studied, for the U{sup 233}-started TMSRs as well as for the Pu-started TMSRs. We will detail optimizations of this transition between the reactors of second and third generations to the Thorium cycle. Such a transition is based on a fleet of TMSRs with no moderator in the core, including TMSRs started with Plutonium and TMSRs directly started with U{sup 233}. We have analyzed the characteristics of each reactor configuration, in terms of deterministic safety parameters, fissile matter inventory, salt reprocessing, radiotoxicity and waste production, and finally deployment capacities.

  9. Spectral structure of electron antineutrinos from nuclear reactors.

    Science.gov (United States)

    Dwyer, D A; Langford, T J

    2015-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape.

  10. SABR fusion-fission hybrid transmutation reactor design concept

    Science.gov (United States)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  11. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    Science.gov (United States)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  12. Results on θ13 Neutrino Oscillations from Reactor Experiments

    Directory of Open Access Journals (Sweden)

    Kim Soo-Bong

    2014-03-01

    Full Text Available Definitive measurements of the smallest neutrino mixing angle θ13 were made by Daya Bay, Double Chooz and RENO in 2012, based on the disappearance of electron antineutrinos emitted from reactors. The new generation reactor experiments have significantly improved a sensitivity for θ13 down to the sin2(2θ13~0.01 level using two identical detectors of 10 ~ 40 tons at near (300 ~ 400 m and far (1 ~ 2 km locations. The θ13 measurements by the three reactor experiments are presented with their future expected sensitivities.

  13. Hygro-Thermo-Mechanical Analysis of a Reactor Vessel

    Directory of Open Access Journals (Sweden)

    Jaroslav Kruis

    2012-01-01

    Full Text Available Determining the durability of a reactor vessel requires a hygro-thermo-mechanical analysis of the vessel throughout its service life. Damage, prestress losses, distribution of heat and moisture and some other quantities are needed for a durability assessment. A coupled analysis was performed on a two-level model because of the huge demands on computer hardware. This paper deals with a hygro-thermo-mechanical analysis of a reactor vessel made of prestressed concrete with a steel inner liner. The reactor vessel is located in Temelín, Czech Republic.

  14. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    CERN Document Server

    Dwyer, D A

    2014-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  15. Radiation effects on reactor pressure vessel supports

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

  16. The Pebble Bed Modular Reactor: An obituary

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

    2011-05-15

    The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

  17. Ageing management experience at NUR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melllal, Sabrina; Rezig, Mohamed; Zamoun, Rachid; Ameur, Azeddin [Nuclear Research Center of Draria, Algiers (Algeria)

    2013-07-01

    NUR is a 1 MW, open pool reactor moderated and cooled by light water. It was commissioned in 1989. NUR is used for education and training in Nuclear Engineering and related topics for COMENA and National Scientific Community. It is also used to perform R and D works and services at national and regional levels. In this presentation, we describe the methodology and the main development activities related to the ageing management at NUR reactor. These activities include inspection actions and development actions to introduce modifications, to solve obsolescence issues in view to implement the required preventive and curative maintenance programs and to improve the performances of the installation. These actions involved mainly the Operation Assistance System of the Reactor (OAS), the secondary cooling loop, the cooling tower. A new OAS using a new technology and having more possibilities than the older one was introduced in the control system of the reactor. The OAS hardware structure, software structure and the main functions performed are presented. The second loop is entirely refurbished. Two new cooling towers are installed and connected to the main heat exchanger with new piping and valves. The architecture of this new installation is described and the performance assessed. Other actions which involve auxiliary systems like emergency electrical system, air pneumatic system and automatic fire extinguishing are presented.

  18. Status and problems of fusion reactor development.

    Science.gov (United States)

    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.

  19. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas

    2013-01-01

    Full Text Available In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA. In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will

  20. New reactors for laboratory studies

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.

    1978-02-01

    Recent developments in design of laboratory and bench-scale reactors reflect mostly the developments in reaction engineering; that is the improved understanding of physical and chemical rate limiting processes, their interactions, and their effects on commercial-scale reactor performance. Whether a laboratory reactor is used to study the fundamentals of a commercial process or for pure scientific interest, it is important to know what physical or chemical process is limiting or influencing the rate and selectivity. To clarify this, a definition is required of the regime where physical influences exist, and study the intrinsic kinetics at conditions where physical processes do not affect the rate. Reactors are illustrated whose design was influenced by the above considerations. These reactors produce results which are independent of the reactors in which they were measured, and which can be scaled up with up-to-date reaction engineering techniques.

  1. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  2. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  3. Acceptability of reactors in space

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  4. Hydrogen Production in Fusion Reactors

    OpenAIRE

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M; Uenosono, C.

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  5. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  6. Space reactor assessment and validation study

    Science.gov (United States)

    Gedeon, Stephen; Morey, Dennis

    The present difficulties experienced by the United States in launching payloads into space has suggested a number of problems which are associated with the handling of hazardous materials in spacecraft. The question has arisen as to the safety of launching highly radioactive material such as plutonium-238, related to the possibility of its dispersion into the atmosphere during a launch vehicle explosion. An alternative is the use of a small nuclear reactor which is not started until it is in space and contains little or no radioactivity at launch. A first order assessment of six small reactor concepts with power levels up to 100 MWe was performed. Both the nuclear feasibility of these concepts to operate at their rated power levels between 7 and 10 years and the capability of these concepts to remain subcritical both before and during launch and also in the case of water immersion during a potential launch failure or abort were investigated.

  7. Final report. U.S. Department of Energy University Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  8. Neutrino Oscillation Studies with Reactors

    CERN Document Server

    Vogel, Petr; Zhang, Chao

    2015-01-01

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  9. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  10. Experimental reactor regulation: the nuclear safety authority's approach; Le controle des reacteurs experimentaux: la demarche de l'Autorite de surete nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Rieu, J.; Conte, D.; Chevalier, A. [Autorite de Surete Nucleaire, 75 - Paris (France)

    2007-07-15

    French research reactors can be classified into 6 categories: 1) critical scale models (Eole, Minerve and Masurca) whose purpose is the study of the neutron production through the fission reaction; 2) reactors that produce neutron beams (Orphee, and the high flux reactor in Grenoble); 3) reactors devoted to safety studies (Cabri, Phebus) whose purpose is to reproduce accidental configurations of power reactors in reduced scale; 4) experimental reactors (Osiris, Phenix) whose purpose is the carrying-out of irradiation experiments concerning nuclear fuels or structure materials; 5) teaching reactors (Ulysse, Isis); and 6) reactors involved in defense programs (Caliban, Prospero, Apareillage-B). We have to note that 3 research reactors are currently being dismantled: Strasbourg University's reactor, Siloe and Siloette. Research reactors in France are of different types and present different hazards. Even if methods of control become more and more similar to those of power reactors, the French Nuclear Safety Authority (ASN) works to allow the necessary flexibility in the ever changing research reactor field while ensuring a high level of safety. Adopting the internal authorizations for operations of minor safety significance, under certain conditions, is one example of this approach. Another challenge in the coming years for ASN is to monitor the ageing of the French research reactors. This includes periodic safety reviews for each facility every ten years. But ASN has also to regulate the new research reactor projects such as Jules Horowitz Reactor, International Thermonuclear Experimental Reactor, which are about to be built.

  11. Accelerator based fusion reactor

    Science.gov (United States)

    Liu, Keh-Fei; Chao, Alexander Wu

    2017-08-01

    A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

  12. Detection of Anomalous Reactor Activity Using Antineutrino Count Rate Evolution Over the Course of a Reactor Cycle

    CERN Document Server

    Bulaevskaya, Vera

    2010-01-01

    This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard 'baseline' fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 73 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be expected by various means, including use of the method i...

  13. Biparticle fluidized bed reactor

    Science.gov (United States)

    Scott, C.D.

    1993-12-14

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  14. FAST NEUTRONIC REACTOR

    Science.gov (United States)

    Snell, A.H.

    1957-12-01

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

  15. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  16. Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2011-05-27

    The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

  17. Small Liquid Metal Cooled Reactor Safety Study

    Energy Technology Data Exchange (ETDEWEB)

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    the core life, which leads to large reactivity worth in the control systems. The conclusions from the evaluations support the high level of safety that can be achieved with small liquid metal cooled reactors using either approach.

  18. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  19. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  20. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  1. Chemical-vapor-deposition reactor

    Science.gov (United States)

    Chern, S.

    1979-01-01

    Reactor utilizes multiple stacked trays compactly arranged in paths of horizontally channeled reactant gas streams. Design allows faster and more efficient deposits of film on substrates, and reduces gas and energy consumption. Lack of dead spots that trap reactive gases reduces reactor purge time.

  2. Thermochemical reactor systems and methods

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-29

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

  3. Test reactor risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  4. Studies on a membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohan, K.; Govind, R.

    1988-10-01

    Simulation is used to evaluate the performance of a catalytic reactor with permeable wall (membrane reactor) in shifting the equilibrium of three reversible reactions (cyclohexane dehydrogenation, hydrogen iodide decomposition, and propylene disproportionation). It is found that the preferred choice of cocurrernt or countercurrent operation is dependent on the physical properties and operating conditions. Methods of enhancing conversion are suggested and temperature effects are discussed.

  5. Thermochemical reactor systems and methods

    Science.gov (United States)

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

    2016-11-29

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

  6. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  7. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  8. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  9. Metallic fuels for advanced reactors

    Science.gov (United States)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  10. A model of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, A.S.; Thompson, B.R.

    1988-09-01

    The analytical model of nuclear reactor transients, incorporating both mechanical and nuclear effects, simulates reactor kinetics. Linear analysis shows the stability borderline for small power perturbations. In a stable system, initial power disturbances die out with time. With an unstable combination of nuclear and mechanical characteristics, initial disturbances persist and may increase with time. With large instability, oscillations of great magnitude occur. Stability requirements set limits on the power density at which particular reactors can operate. The limiting power density depends largely on the product of two terms: the fraction of delayed neutrons and the frictional damping of vibratory motion in reactor core components. As the fraction of delayed neutrons is essentially fixed, mechanical damping largely determines the maximum power density. A computer program, based on the analytical model, calculates and plots reactor power as a nonlinear function of time in response to assigned values of mechanical and nuclear characteristics.

  11. Engineering reactors for catalytic reactions

    Indian Academy of Sciences (India)

    Vivek V Ranade

    2014-03-01

    Catalytic reactions are ubiquitous in chemical and allied industries. A homogeneous or heterogeneous catalyst which provides an alternative route of reaction with lower activation energy and better control on selectivity can make substantial impact on process viability and economics. Extensive studies have been conducted to establish sound basis for design and engineering of reactors for practising such catalytic reactions and for realizing improvements in reactor performance. In this article, application of recent (and not so recent) developments in engineering reactors for catalytic reactions is discussed. Some examples where performance enhancement was realized by catalyst design, appropriate choice of reactor, better injection and dispersion strategies and recent advances in process intensification/ multifunctional reactors are discussed to illustrate the approach.

  12. Unsteady processes in catalytic reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matros, Yu.Sh.

    1985-01-01

    In recent years a realization has occurred that reaction and reactor dynamics must be considered when designing and operating catalytic reactors. In this book, the author has focussed on both the processes occurring on individual porous-catalyst particles as well as the phenomena displayed by collections of these particles in fixed-bed reactors. The major topics discussed include the effects of unsteady-state heat and mass transfer, the influence of inhomogeneities and stagnant regions in fixed beds, and reactor operation during forced cycling of operating conditions. Despite the title of the book, attention is also paid to the determination of the number and stability of fixed-bed steady states, with the aim of describing the possibility of controlling reactors at unstable steady states. However, this development is somewhat dated, given the recent literature on multiplicity phenomena and process control.

  13. Antineutrino Monitoring of Thorium Reactors

    CERN Document Server

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  14. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  15. WILDCAT: a catalyzed D-D tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

    1981-11-01

    WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.

  16. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

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

    2016-01-01

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

  17. A Study on the Kinetic Characteristics of Transmutation Process Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae seon; Huh, Chang Wook; Kim, Doh Hyung [Seoul National University, Seoul (Korea, Republic of)

    1997-07-01

    The purpose of this study is to examine the transient heat transfer characteristics of liquid mental as the coolant used in accelerator-driven transmutation process reactor which is related the disposal of high-level radioactive nuclide. At current stage, the accelerator-driven transmutation process is investigated as the most appropriate method among many transmutation process methods. In this study, previous research works are investigated especially about the thermal hydraulics and kinetic behavior of coolant material including heat transfer of coolant in transmutation process reactor. A study on the heat transfer characteristics of liquid metal is performed based on the thermal hydraulic kinetic characteristics of liquid metal reactor which uses liquid metal coolant. Based on this study, the most appropriate material for the coolant of transmutation reactor will be recommended. 53 refs., 15 tabs., 33 figs. (author)

  18. Studies on Nitrogen Oxides Removal Using Plasma Assisted Catalytic Reactor

    Institute of Scientific and Technical Information of China (English)

    V. Ravi; Young Sun Mok; B. S. Rajanikanth; Ho-Chul Kang

    2003-01-01

    An electric discharge plasma reactor combined with a catalytic reactor was studied for removing nitrogen oxides. To understand the combined process thoroughly, discharge plasma and catalytic process were separately studied first, and then the two processes were combined for the study. The plasma reactor was able to oxidize NO to NO2 well although the oxidation rate decreased with temperature. The plasma reactor alone did not reduce the NOx (NO+NO2)level effectively, but the increase in the ratio of NO2 to NO as a result of plasma discharge led to the enhancement of NOx removal efficiency even at lower temperatures over the catalyst surface (V2O5-WOa/TiO2). At a gas temperature of 100℃, the NOx removal efficiency obtained using the combined plasma catalytic process was 88% for an energy input of 36 eV/molecule or 30 J/1.

  19. Background radiation measurements at high power research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ashenfelter, J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Balantekin, B. [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Baldenegro, C.X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Band, H.R. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Barclay, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Bass, C.D. [Department of Chemistry and Physics, Le Moyne College, Syracuse, NY 13214 (United States); Berish, D. [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Bowden, N.S., E-mail: nbowden@llnl.gov [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bryan, C.D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherwinka, J.J. [Physical Sciences Laboratory, University of Wisconsin, Madison, WI 53706 (United States); Chu, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Classen, T. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Davee, D. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Dean, D.; Deichert, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Dolinski, M.J. [Department of Physics, Drexel University, Philadelphia, PA 19104 (United States); Dolph, J. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Dwyer, D.A. [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Fan, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); and others

    2016-01-11

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

  20. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  1. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  2. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  3. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  4. Neutronic Reactor Shield

    Science.gov (United States)

    Fermi, Enrico; Zinn, Walter H.

    The argument of the present Patent is a radiation shield suitable for protection of personnel from both gamma rays and neutrons. Such a shield from dangerous radiations is achieved to the best by the combined action of a neutron slowing material (a moderator) and a neutron absorbing material. Hydrogen is particularly effective for this shield since it is a good absorber of slow neutrons and a good moderator of fast neutrons. The neutrons slowed down by hydrogen may, then, be absorbed by other materials such as boron, cadmium, gadolinium, samarium or steel. Steel is particularly convenient for the purpose, given its effectiveness in absorbing also the gamma rays from the reactor (both primary gamma rays and secondary ones produced by the moderation of neutrons). In particular, in the present Patent a shield is described, made of alternate layers of steel and Masonite (an hydrolized ligno-cellulose material). The object of the present Patent is not discussed in any other published paper.

  5. Neutrino Experiments at Reactors

    Science.gov (United States)

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  6. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  7. The application of research reactor Maria for analysis of thorium use in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S.; Andrzejewski, K.; Myslek-Laurikainen, B.; Pytel, B.; Szczurek, J. [Dep. Thorium Project, Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk (Poland); Polkowska-Motrenko, H. [Institute of Nuclear Chemistry and Technology, ul.Dorodna 16 03-195 Warszawa (Poland)

    2010-07-01

    The MARIA reactor, pool-type light-water cooled and beryllium moderated nuclear research reactor was used to evaluate the {sup 233}U breeding during the experimental irradiation of the thorium samples. The level of impurities concentrations was determined using ICP-MS method. The associated development of computer programs for analysis of application of thorium in EPR reactor consist of PC version of CORD-2/GNOMER system are presented. (authors)

  8. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  9. That Great Leviathan - or: how to move a reactor from Cumbria to Caithness

    Energy Technology Data Exchange (ETDEWEB)

    The paper concerns the moving of a complete nuclear reactor from Barrow-in-Furness to Dounreay, Scotland. The nuclear reactor is the PWR2, which is the prototype for the new reactors that will propel larger submarines with low noise levels that are difficult for an enemy to detect. The shipment of the PWR2 and its auxiliary equipment, but without any fissile material, is briefly described. (U.K.).

  10. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  11. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  12. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  13. Reactor neutrons in nuclear astrophysics

    Science.gov (United States)

    Reifarth, René; Glorius, Jan; Göbel, Kathrin; Heftrich, Tanja; Jentschel, Michael; Jurado, Beatriz; Käppeler, Franz; Köster, Ulli; Langer, Christoph; Litvinov, Yuri A.; Weigand, Mario

    2017-09-01

    The huge neutron fluxes offer the possibility to use research reactors to produce isotopes of interest, which can be investigated afterwards. An example is the half-lives of long-lived isotopes like 129I. A direct usage of reactor neutrons in the astrophysical energy regime is only possible, if the corresponding ions are not at rest in the laboratory frame. The combination of an ion storage ring with a reactor and a neutron guide could open the path to direct measurements of neutron-induced cross sections on short-lived radioactive isotopes in the astrophysically interesting energy regime.

  14. Nuclear reactor downcomer flow deflector

    Energy Technology Data Exchange (ETDEWEB)

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15

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

  15. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  16. Analysis of Reactor Deployment Scenarios with Introduction of SFR Breakeven Reactors and Burners Using DANESS Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2008-01-15

    Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.

  17. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  18. Breeder Reactors, Understanding the Atom Series.

    Science.gov (United States)

    Mitchell, Walter, III; Turner, Stanley E.

    The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

  19. Evolution of the tandem mirror reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, G.A.; Logan, B.G.

    1982-03-09

    We discuss the evolution of the tandem mirror reactor concept from the original conceptual reactor design (1977) through the first application of the thermal barrier concept to a reactor design (1979) to the beginning of the Mirror Advanced Reactor Study (1982).

  20. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  1. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented.

  2. Microchannel Reactor System for Catalytic Hydrogenation

    Energy Technology Data Exchange (ETDEWEB)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  3. NRC policy on future reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-07-01

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  4. Advanced Nuclear Reactor Concepts for China

    Energy Technology Data Exchange (ETDEWEB)

    Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

    2008-07-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  5. Proliferation resistance of small modular reactors fuels

    Energy Technology Data Exchange (ETDEWEB)

    Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

    2013-07-01

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  6. NRC policy on future reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-07-01

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  7. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  8. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  9. Site Investigation for Detection of KIJANG Reactor Core Center

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae-Hyun; Kim, Jun Yeon; Kim, Jeeyoung [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    It was planned for the end of March 2017 and extended to April 2018 according to the government budget adjustment. The KJRR project is intended for filling the self-sufficiency of RI demand including Mo-99, increasing the NTD capacity and developing technologies related to the research reactor. In project, site investigation is the first activity that defines seismologic and related geologic aspects of the site. Site investigation was carried out from Oct. 2012 to Jan. 2014 and this study is intended to describe detail procedures in locating the reactor core center. The location of the reactor core center was determined by collectively reviewing not only geological information but also information from architects engineering. EL 50m was selected as ground level by levering construction cost. Four recommended locations (R-1a - R-1d) are displayed for the reactor core center. R-1a was found optimal in consideration of medium rock contour, portion of medium rock covering reactor buildings, construction cost, physical protection and electrical resistivity. It is noted that engineering properties of the medium rock is TCR/RQD 100/53, elastic modulus 7,710 - 8,720MPa, permeability coefficient 2.92E-06cm/s, and S-wave velocity 1,380m/s, sound for foundations of reactor buildings.

  10. Conceptual Study for development of a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, C.; Kim, H. S.; Park, J. H.; Chae, H. T.; Lee, B. C. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    Even though the nuclear society is again facing with difficult situations after Fukusima accident, some countries still continues to consider nuclear power as one option of national energy sources and to introduce nuclear energy. As a research reactor has been regarded as a step-stone to establish infrastructures for the nuclear power development program, some countries that have plan to introduce the nuclear power energy are considering to construct a research reactor. Particularly, a low power research reactor whose main purpose is basic researches on the nuclear technology and education/training would be of interest to developing countries when taking the economy and level of science and technology into consideration. And many low power research reactors at operation are obsolescent and their numbers are decreasing. Hence, some concepts on a low power research reactor are being studied for the future needs. This paper presents the conceptual study on the basic requirements and the preliminary design features of a low power research reactor.

  11. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  12. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF/sub 6/ and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 233/U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of /sup 233/U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors.

  13. Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Rubindamayugi, M.S.T.; Salakana, L.K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

    1997-12-31

    The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

  14. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  15. Thermal Analysis for Mobile Reactor

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>Mobile reactor design in the paper is consisted of two grades of thermal electric conversion. The first grade is the thermionic conversion inside the core and the second grade is thermocouple conversion

  16. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  17. Teaching About Nature's Nuclear Reactors

    CERN Document Server

    Herndon, J M

    2005-01-01

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  18. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  19. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

    1989-12-01

    This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

  20. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  1. Unique features of space reactors

    Science.gov (United States)

    Buden, David

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K.

  2. Small reactor power systems for manned planetary surface bases

    Science.gov (United States)

    Bloomfield, Harvey S.

    1987-01-01

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

  3. Radioactive waste management and disposal scenario for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

    1997-10-01

    The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

  4. Reactor antineutrinos and nuclear physics

    Science.gov (United States)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  5. Microchannel Reactors for ISRU Applications

    Science.gov (United States)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  6. The resonance absorption controlled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R.

    1977-07-01

    In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D{sub 2}O/H{sub 2}O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs.

  7. Calculation of reactor antineutrino spectra in TEXONO

    CERN Document Server

    Chen Dong Liang; Mao Ze Pu; Wong, T H

    2002-01-01

    In the low energy reactor antineutrino physics experiments, either for the researches of antineutrino oscillation and antineutrino reactions, or for the measurement of abnormal magnetic moment of antineutrino, the flux and the spectra of reactor antineutrino must be described accurately. The method of calculation of reactor antineutrino spectra was discussed in detail. Furthermore, based on the actual circumstances of NP2 reactors and the arrangement of detectors, the flux and the spectra of reactor antineutrino in TEXONO were worked out

  8. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  9. Novel Catalytic Membrane Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stuart Nemser, PhD

    2010-10-01

    There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

  10. Tritium management in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II.

  11. Utilization of Minor Actinides (Np, Am, Cm) in Nuclear Power Reactor

    Science.gov (United States)

    Gerasimov, A.; Bergelson, B.; Tikhomirov, G.

    2014-06-01

    Calculation research of the utilization process of minor actinides (transmutation with use of power released) is performed for specialized power reactor of the VVER type operating on the level of electric power of 1000 MW. Five subsequent cycles are considered for the reactor with fuel elements containing minor actinides along with enriched uranium. It was shown that one specialized reactor for the one cycle (900 days) can utilize minor actinides from several VVER-1000 reactors without any technological and structural modifications. Power released because of minor actinide fission is about 4% with respect to the total power

  12. Simultaneous hydrogen utilization and in situ biogas upgrading in an anaerobic reactor

    DEFF Research Database (Denmark)

    Luo, Gang; Johansson, Sara; Boe, Kanokwan;

    2011-01-01

    . The methane production rate of the reactor with H2 addition was 22% higher, compared to the control reactor only fed with manure. The CO2 content in the produced biogas was only 15%, while it was 38% in the control reactor. However, the addition of hydrogen resulted in increase of pH (from 8.0 to 8.3) due......The possibility of converting hydrogen to methane and simultaneous upgrading of biogas was investigated in both batch tests and fully mixed biogas reactor, simultaneously fed with manure and hydrogen. Batch experiments showed that hydrogen could be converted to methane by hydrogenotrophic...... mixing intensity (shaking speed 300 rpm). Continuous addition of hydrogen (flow rate of 28.6 mL/(L/h)) to an anaerobic reactor fed with manure, showed that more than 80% of the hydrogen was utilized. The propionate and butyrate level in the reactor was not significantly affected by the hydrogen addition...

  13. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

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

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2008-06-01

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

  15. An assessment of space reactor technology needs and recommendations for development

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C. [Sandia National Labs., Albuquerque, NM (United States); Wiley, R.L. [Consultant, Columbia, MD (United States)

    1995-11-01

    In order to provide a strategy for space reactor technology development, the Defense Nuclear Agency (DNA) has authorized a brief review of potential national needs that may be addressed by space reactor systems. a systematic approach was used to explore needs at several levels that are increasingly specific. Level 0 -- general trends and issues; Level 1 -- generic space capabilities to address trends; Level 2 -- requirements to support capabilities; Level 3 -- system types capable of meeting requirements; Level 4 --generic reactor system types; and Level 5 -- specific baseline systems. Using these findings, a strategy was developed to support important space reactor technologies within a limited budget. A preliminary evaluation identified key technical issues and provide a prioritized set of candidate research projects. The evaluation of issues and the recommended research projects are presented in a companion paper.

  16. Selection of nuclear reactors through the hierarchic analysis process: the Mexican case; Seleccion de reactores nucleares mediante el proceso de analisis jerarquico: el caso Mexicano

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2008-07-01

    In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best

  17. Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

    Energy Technology Data Exchange (ETDEWEB)

    Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

    2011-07-01

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

  18. Establishment of licensing process for development reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik (and others)

    2006-02-15

    A study on licensing processes for development reactors has been performed to prepare the licensing of development reactors developed in Korea. The contents and results of the study are summarized as follows. The licensing processes for nuclear reactors in Korea, U.S.A., Japan, France, U.K., Canada, and IAEA were surveyed and analyzed to obtain technical bases necessary for establishing licensing processes applicable to development reactors in Korea. Based on the technical bases obtained the above analysis, the purpose, power output, and design characteristics of development reactors were analyzed in detail. The analysis results suggested that development reactors should be classified as a new reactor category (called as 'development reactor') separated from the current reactor categories such as the research reactor and the power reactor. Therefore, it is proposed to establish a new reactor category classified as 'development reactor' for the development reactors. And licensing processes, including licensing technical requirements, licensing document requirements, and other regulatory requirements, were also proposed for the development reactors. In order to institutionalize the licensing processes developed in this study, it is necessary to revise the current laws. Therefore, draft provisions of Atomic Energy Act, Enforcement Decree of the Atomic Energy Act, and Enforcement Regulation of the Atomic Energy Act have been developed for the preparation of the future legalization of the licensing processes proposed for the development reactors. Conclusively, a proposal of licensing processes and draft provisions of laws have been developed for the development reactors. The results proposed in this study can be applied directly to the licensing of the future development reactors. Furthermore, they will also contribute to establishing successfully the licensing processes of the development reactors.

  19. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

  20. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

  1. Reactivity determination in accelerator driven reactors using reactor noise analysis

    Directory of Open Access Journals (Sweden)

    Kostić Ljiljana 1

    2002-01-01

    Full Text Available Feynman-alpha and Rossi-alpha methods are used in traditional nuclear reactors to determine the subcritical reactivity of a system. The methods are based on the measurement of the mean value, variance and the covariance of detector counts for different measurement times. Such methods attracted renewed attention recently with the advent of the so-called accelerator driven reactors (ADS proposed some time ago. The ADS systems, intended to be used either in energy generation or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those traditionally used by radioactive sources. In such reactors the monitoring of the subcritical reactivity is very important, and a statistical method, such as the Feynman-alpha method, is capable of resolving this problem.

  2. Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

    2009-10-15

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  3. Present state of the liner of the reactor; Estado actual del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Raya A, R.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    When being presented to work the operation personnel of the reactor, on Monday January 10, 1983, they noticed that the reactor pool was overflowing of water and the floor of the room was partially flooded. The personnel proceeded to revise the feedwater systems to the pool, the Emergency Cooling System of the core and that of Water of Reinstatement, was found that the passing valve of this last it was lightly open. It was discovered that the water that was flooded in the floor of the room it came from the relief valves of the ports TW-1 and RW-2 and of three glides that were in the Thermal Column area. It was proceeded to lower the one level of water of the pool to their normal position and it was clean the water flooded in the salts. (Author)

  4. Thermonuclear Reflect AB-Reactor

    CERN Document Server

    Bolonkin, Alexander

    2008-01-01

    The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical pr...

  5. Heterogeneous Transmutation Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  6. Entropy Production in Chemical Reactors

    Science.gov (United States)

    Kingston, Diego; Razzitte, Adrián C.

    2017-06-01

    We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

  7. Simplifying Microbial Electrosynthesis Reactor Design

    Directory of Open Access Journals (Sweden)

    Cloelle G.S. Giddings

    2015-05-01

    Full Text Available Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata, which reduces carbon dioxide to acetate. In traditional ‘H-cell’ reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a poteniostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs.

  8. Hanford reactor and separations facility advantages

    Energy Technology Data Exchange (ETDEWEB)

    1963-06-27

    This document describes the advantages and limitations of Hanford production facilities. In addition to summarizing the technical parameters of the reactors and separations plants and their mechanical features, the unique aspects of these facilities to the production of special materials in which the Commission may be interested have been discussed. As the primary difference between the B-C-D-DR-F-H reactors and the K reactors and the K reactors is in the number and length of process channels. This report is addressed primarily to the 2000-tube reactors. K reactor characteristics are within the range of lattice and flexibility parameters described.

  9. Imaging Fukushima Daiichi reactors with muons

    Directory of Open Access Journals (Sweden)

    Haruo Miyadera

    2013-05-01

    Full Text Available A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  10. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  11. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  12. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  13. Mixing In Jet-Stirred Reactors With Different Geometries

    KAUST Repository

    Ayass, Wassim W.

    2013-12-01

    This work offers a well-developed understanding of the mixing process inside Jet- Stirred Reactors (JSR’s) with different geometries. Due to the difficulty of manufacturing these JSR’s made in quartz, existing JSR configurations were assessed with certain modifications and optimal operating conditions were suggested for each reactor. The effect of changing the reactor volume, the nozzle diameter and shape on mixing were both studied. Two nozzle geometries were examined in this study, a crossed shape nozzle and an inclined shape nozzle. Overall, six reactor configurations were assessed by conducting tracer experiments - using the state-of-art technologies of high-speed cameras and laser absorption spectroscopy- and Computational Fluid Dynamics (CFD) simulations. The high-speed camera tracer experiment gives unique qualitative information – not present in the literature – about the actual flow field. On the other hand, when using the laser technique, a more quantitative analysis emerges with determining the experimental residence time distribution (RTD) curves of each reactor. Comparing these RTD curves with the ideal curve helped in eliminating two cases. Finally, the CFD simulations predict the RTD curves as well as the mixing levels of the JSR’s operated at different residence times. All of these performed studies suggested the use of an inclined nozzle configuration with a reactor diameter D of 40mm and a nozzle diameter d of 1mm as the optimal choice for low residence time operation. However, for higher residence times, the crossed configuration reactor with D=56mm and d=0.3mm gave a nearly perfect behavior.

  14. Realtime control of biogas reactors. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Poulsen, Allan K.

    2010-12-15

    . However, when having a periodic returning gas flow across the membrane for several months components in the gas accumulated in the MIMS which had to be opened and cleaned periodically. The MIMS also had to be calibrated at regular intervals and is a much more laborious method compared to {mu}-GC for a biogas environment. 3) VFA automaton is in theory a very promising tool for real time control of biogas reactors. However, the technique was too unreliable in praxis and the technology still needs to be further developed. 4) Raman spectroscopy preliminary studies carried out in this project showed that no dedicated peaks could be found in the Raman spectra. Raman spectroscopy is however a promising technology that is fast, reliable and still becoming cheaper and cheaper for purchasing. Since no dedicated peaks were found a much more detailed study will need to be carried out, also in combination with chemometric analyses of the spectra which were out of the scope in this project. 5) NIRS together with chemometric analyses of the spectra and VFA levels determined by manual based sampling and quantification by GC has in scientific papers shown promising for online quantification of VFAs. However, the conclusion in this project is that the quality and reliability of NIR models is low when attempting to predict VFAs in real time in a complex media such as biogas. (LN)

  15. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.

  16. Operational transparency: an advanced safeguards strategy for future on-load refuelled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Whitlock, J.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Trask, D. [Atomic Energy of Canada Limited, Fredericton, New Brunswick (Canada)

    2012-03-15

    The IAEA's system for tracking fuel movement in an on-load refuelled heavy-water reactor is robust, but an opportunity remains to exploit the wealth of data streaming from the reactor vault during operation and provide real-time, third-party monitoring of reactor status and history. This concept of Operational Transparency would require that large amounts of operational data be reduced in near-real time to a small subset of high-level information. Operational Transparency would enhance the IAEA's ability to monitor the state of the core to an unprecedented level. This paper provides an overview of the novel concept of Operational Transparency in heavy water reactors, using potential application to CANDU reactors as an example, and explores some of the technical challenges that will need to be solved for efficient implementation. (author)

  17. A tubular focused sonochemistry reactor

    Institute of Scientific and Technical Information of China (English)

    ZHOU GuangPing; LIANG ZhaoFeng; LI ZhengZhong; ZHANG YiHui

    2007-01-01

    This paper presents a new sonochemistry reactor, which consists of a cylindrical tube with a certain length and piezoelectric transducers at tube's end with the longitudinal vibration. The tube can effectively transform the longitudinal vibration into the radial vibration and thereby generates ultrasound. Furthermore, ultrasound can be focused to form high-intensity ultrasonic field inside tube. The reactor boasts of simple structure and its whole vessel wall can radiate ultrasound so that the electroacoustic transfer efficiency is high. The focused ultrasonic field provides good condition for sonochemical reaction. The length of the reactor can be up to 2 meters, and liquids can pass through it continuously, so it can be widely applied in liquid processing such as sonochemistry.

  18. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

    The low aspect ration tokamak is proposed for the driver of a transmutation reactor.The main parameters of the reactor core,neutronic analysis of the blanket are given>the neutron wall loading can be lowered from the magnitude order of 1 MW/m2 to 0.5MW/m2 which is much easier to reach in the near future,and the transmutation efficiency (fission/absorption ratio)is raised further.The blanket power density is about 200MW/m3 which is not difficult to deal with.The key components such as diverter and center conductor post are also designed and compared with conventional TOkamak,Finally,by comparison with the other drivers such as FBR,PWR and accelerator,it can be anticipated that the low aspect ratio transmutation reactor would be one way of fusion energy applications in the near future.

  19. Plasma reactor waste management systems

    Science.gov (United States)

    Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

    1992-01-01

    The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

  20. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  1. Utilisation of thorium in reactors

    Science.gov (United States)

    Anantharaman, K.; Shivakumar, V.; Saha, D.

    2008-12-01

    India's nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th-Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle.

  2. Novel Technology for Phenol Wastewater Treatment Using Electrochemical Reactor

    Directory of Open Access Journals (Sweden)

    Yuncheng Xie

    2015-01-01

    Full Text Available There are various electrochemical approaches to save energy, mostly by means of equipment improvement coupled with other water treatment technologies. Replacement of DC power with pulse power, modified reactor coupled with photocatalysis can decrease cost. But more or less additional input is developed, or infrastructure has to be replaced. In this paper, an N-Step electrochemical reactor, based on stage reaction modeling, is put forward. On the basis of not changing equipment investment and by adjustment of the operating current density at different levels, power consumption decreases. This model develops a foundation of electrochemical water treatment technology for the engineering application.

  3. An accelerator-driven reactor for meeting future energy demand

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroshi; Yang, Y.; Yu, A.

    1997-12-31

    Fissile fuel can be produced at a high rate using an accelerator-driven Pu-fueled subcritical fast reactor which avoids encountering a shortage of Pu during a high growth rate in the production of nuclear energy. Furthermore, the necessity of the early introduction of the fast reactor can be moderated. Subcritical operation provides flexible nuclear energy options along with high neutron economy for producing the fuel, for transmuting high-level waste such as minor actinides, and for efficiently converting excess and military Pu into proliferation-resistant fuel.

  4. External fuel thermionic reactor system.

    Science.gov (United States)

    Mondt, J. F.; Peelgren, M. L.

    1971-01-01

    Thermionic reactors are prime candidates for nuclear electric propulsion. The national thermionic reactor effort is concentrated on the flashlight concept with the external-fuel concept as the backup. The external-fuel concept is very adaptable to a completely modular power subsystem which is attractive for highly reliable long-life applications. The 20- to 25-cm long, externally-fueled converters have been designed, fabricated, and successfully tested with many thermal cycles by electrical heating. However, difficulties have been encountered during encapsulation for nuclear heated tests and none have been started to date. These nuclear tests are required to demonstrate the concept feasibility.

  5. Analysis of Adiabatic Batch Reactor

    Directory of Open Access Journals (Sweden)

    Erald Gjonaj

    2016-05-01

    Full Text Available A mixture of acetic anhydride is reacted with excess water in an adiabatic batch reactor to form an exothermic reaction. The concentration of acetic anhydride and the temperature inside the adiabatic batch reactor are calculated with an initial temperature of 20°C, an initial temperature of 30°C, and with a cooling jacket maintaining the temperature at a constant of 20°C. The graphs of the three different scenarios show that the highest temperatures will cause the reaction to occur faster.

  6. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  7. Reactor shutdown delays medical procedures

    Science.gov (United States)

    Gwynne, Peter

    2008-01-01

    A longer-than-expected maintenance shutdown of the Canadian nuclear reactor that produces North America's entire supply of molybdenum-99 - from which the radioactive isotopes technetium-99 and iodine-131 are made - caused delays to the diagnosis and treatment of thousands of seriously ill patients last month. Technetium-99 is a key component of nuclear-medicine scans, while iodine-131 is used to treat cancer and other diseases of the thyroid. Production eventually resumed, but only after the Canadian government had overruled the Canadian Nuclear Safety Commission (CNSC), which was still concerned about the reactor's safety.

  8. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  9. Scaledown of a methanol reactor

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.

    1983-07-01

    This article shows how it is possible to define operating conditions for pilot plants and development labs by scaling down a commercial reactor. Points out that scaledown consideration and experiment planning can be done in a similar manner for the boiling water-cooled, Lurgi-type reactor. Explains that although the design of large, single-train plants to produce methanol for fuel use has different economic objectives, product specifications, and technical constraints from the traditional commercial methanol plants, the same fundamental laws of thermodynamics and reaction kinetics apply to both types of operation.

  10. Neutron flux optimization in irradiation channels at NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meftah, B. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria)]. E-mail: b_meftah@yahoo.com; Zidi, T. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria); Bousbia-Salah, A. [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2 - 56126 Pisa (Italy)

    2006-09-15

    Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

  11. Process Intensification via Membrane Reactors, the DEMCAMER Project

    Directory of Open Access Journals (Sweden)

    Fausto Gallucci

    2016-05-01

    Full Text Available This paper reports the findings of a FP7 project (DEMCAMER that developed materials (catalysts and membranes and new processes for four industrially relevant reaction processes. In this project, active, stable, and selective catalysts were developed for the reaction systems of interest and their production scaled up to kg scale (TRL5 (TRL: Technology Readiness Level. Simultaneously, new membranes for gas separation were developed; in particular, dense supported thin palladium-based membranes for hydrogen separation from reactive mixtures. These membranes were successfully scaled up to TRL4 and used in various lab-scale reactors for water gas shift (WGS, using both packed bed and fluidized bed reactors, and Fischer-Tropsch (FTS using packed bed reactors and in prototype reactors for WGS and FTS. Mixed ionic-electronic conducting membranes in capillary form were also developed for high temperature oxygen separation from air. These membranes can be used for both Autothermal Reforming (ATR and Oxidative Coupling of Methane (OCM reaction systems to increase the efficiency and the yield of the processes. The production of these membranes was scaled up to TRL3–4. The project also developed adequate sealing techniques to be able to integrate the different membranes in lab-scale and prototype reactors.

  12. Perturbation analysis of the TRIGA Mark II reactor Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)

    2013-04-15

    The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

  13. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  14. Effect of DUPIC Cycle on CANDU Reactor Safety Parameters

    Directory of Open Access Journals (Sweden)

    Nader M.A. Mohamed

    2016-10-01

    Full Text Available Although, the direct use of spent pressurized water reactor (PWR fuel in CANda Deuterium Uranium (CANDU reactors (DUPIC cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO2 enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1 the power distribution amongst the fuel elements of the bundle; (2 the coolant void reactivity; and (3 the reactor point-kinetics parameters.

  15. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

  16. A probabilistic consequence assessment for a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2017-02-15

    Currently, fossil fuel is globally running out. If current trends continue, crude oil will be depleted in 20 years and natural gas in 40 years. In addition, the use of fossil resource has increased emissions of green gas such as carbon dioxide. Therefore, there has been a strong demand in recent years for producing large amounts of hydrogen as an alternative energy [1]. To generate hydrogen energy, very high temperature more than 900 C is required but this level is not easy to reach. Because a Very High Temperature Reactor (VHTR), one of next generation reactor, is able to make the temperature, it is regarded as a solution of the problem. Also, VHTR has an excellent safety in comparison with existing and other next generation reactors. Especially, a passive system, Reactor Cavity Cooling System (RCCS), is adopted to get rid of radiant heat in case of accidents. To achieve variety requirements of new designed-reactors, however, it needs to develop new methodologies and definitions different with existing method. At the same time, an application of probability safety assessment (PSA) has been proposed to ensure the safety of next generation NPPs. For this, risk-informed designs of structures have to be developed and verified. Particularly, the passive system requires to be evaluated for its reliability. The objective of this study is to improve safety of VIITR by conducting risk profile.

  17. A novel concept for CRIEC-driven subcritical research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, M.; Miley, G.H. [Illinois Univ., Fusion Studies Lab., Dept. of Nuclear, Plasma, and Radiological Engineering, Urbana, IL (United States)

    2001-07-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  18. Piezoelectric material for use in a nuclear reactor core

    Science.gov (United States)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

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

    Science.gov (United States)

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

    2012-06-01

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

  20. COMSORS: A light water reactor chemical core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.; Kenton, M.A. [Creare Inc., Hanover, NH (United States)

    1997-02-24

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation.

  1. Reactor Antineutrino Signals at Morton and Boulby

    CERN Document Server

    Dye, Steve

    2016-01-01

    Increasing the distance from which an antineutrino detector is capable of monitoring the operation of a registered reactor, or discovering a clandestine reactor, strengthens the Non-Proliferation of Nuclear Weapons Treaty. This report presents calculations of reactor antineutrino interactions, from quasi-elastic neutrino-proton scattering and elastic neutrino-electron scattering, in a water-based detector operated >10 km from a commercial power reactor. It separately calculates signal from the proximal reactor and background from all other registered reactors. The main results are interaction rates and kinetic energy distributions of charged leptons scattered from quasi-elastic and elastic processes. Comparing signal and background distributions evaluates reactor monitoring capability. Scaling the results to detectors of different sizes, target media, and standoff distances is straightforward. Calculations are for two examples of a commercial reactor (P_th~3 GW) operating nearby (L~20 km) an underground facil...

  2. Oregon State University TRIGA Reactor annual report

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

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

  4. Segmentation of the internal of the Reactor Jose Cabrera NPP; Segmentacion de los Internos del Reactor CN Jose Cabrera

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez silva, M.; Borque Linan, J.

    2013-07-01

    The Plan of dismantling and decommissioning of the Jose Cabrera NPP represents the first total dismantling of a nuclear power station in Spain (level 3 of the IAEA). Complete disassembly of the different components of the primary circuit (internal reactor vessel, pusher, generator of steam, etc.) represents a differential activity against previous projects of dismantling The segmentation of the inmates of the reactor under water using tele operators cutting tools in the spent fuel pit, has been a challenge from the technological point of view as well as a critical activity in the framework of the radiological dis-assemblies associated to the Plan of dismantling and Decommissioning of the Jose Cabrera NPP.

  5. Transmutation of actinides in power reactors.

    Science.gov (United States)

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

    2005-01-01

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

  6. Nitrification of industrial and domestic saline wastewaters in moving bed biofilm reactor and sequencing batch reactor.

    Science.gov (United States)

    Bassin, João P; Dezotti, Marcia; Sant'anna, Geraldo L

    2011-01-15

    Nitrification of saline wastewaters was investigated in bench-scale moving-bed biofilm reactors (MBBR). Wastewater from a chemical industry and domestic sewage, both treated by the activated sludge process, were fed to moving-bed reactors. The industrial wastewater contained 8000 mg Cl(-)/L and the salinity of the treated sewage was gradually increased until that level. Residual substances present in the treated industrial wastewater had a strong inhibitory effect on the nitrification process. Assays to determine inhibitory effects were performed with the industrial wastewater, which was submitted to ozonation and carbon adsorption pretreatments. The latter treatment was effective for dissolved organic carbon (DOC) removal and improved nitrification efficiency. Nitrification percentage of the treated domestic sewage was higher than 90% for all tested chloride concentrations up to 8000 mg/L. Results obtained in a sequencing batch reactor (SBR) were consistent with those attained in the MBBR systems, allowing tertiary nitrification and providing adequate conditions for adaptation of nitrifying microorganisms even under stressing and inhibitory conditions.

  7. Nitrification of industrial and domestic saline wastewaters in moving bed biofilm reactor and sequencing batch reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bassin, Joao P. [Programa de Engenharia Quimica/COPPE, Universidade Federal do Rio de Janeiro, Centro de Tecnologia, Bloco G - sala 116, P.O. Box 68502, 21941-972 Rio de Janeiro, RJ (Brazil); Dezotti, Marcia, E-mail: mdezotti@peq.coppe.ufrj.br [Programa de Engenharia Quimica/COPPE, Universidade Federal do Rio de Janeiro, Centro de Tecnologia, Bloco G - sala 116, P.O. Box 68502, 21941-972 Rio de Janeiro, RJ (Brazil); Sant' Anna, Geraldo L. [Programa de Engenharia Quimica/COPPE, Universidade Federal do Rio de Janeiro, Centro de Tecnologia, Bloco G - sala 116, P.O. Box 68502, 21941-972 Rio de Janeiro, RJ (Brazil)

    2011-01-15

    Nitrification of saline wastewaters was investigated in bench-scale moving-bed biofilm reactors (MBBR). Wastewater from a chemical industry and domestic sewage, both treated by the activated sludge process, were fed to moving-bed reactors. The industrial wastewater contained 8000 mg Cl{sup -}/L and the salinity of the treated sewage was gradually increased until that level. Residual substances present in the treated industrial wastewater had a strong inhibitory effect on the nitrification process. Assays to determine inhibitory effects were performed with the industrial wastewater, which was submitted to ozonation and carbon adsorption pretreatments. The latter treatment was effective for dissolved organic carbon (DOC) removal and improved nitrification efficiency. Nitrification percentage of the treated domestic sewage was higher than 90% for all tested chloride concentrations up to 8000 mg/L. Results obtained in a sequencing batch reactor (SBR) were consistent with those attained in the MBBR systems, allowing tertiary nitrification and providing adequate conditions for adaptation of nitrifying microorganisms even under stressing and inhibitory conditions.

  8. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  9. Kinetic Parameter Measurements in the MINERVE Reactor

    Science.gov (United States)

    Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

    2017-01-01

    In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

  10. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)

    2003-07-01

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  11. Heavy Water Reactor; Reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)

    2000-04-01

    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  12. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  13. Some new viewpoints in reactor noise analysis

    Institute of Scientific and Technical Information of China (English)

    罗征培; 李富; 等

    1996-01-01

    It is propsed that the linearity criterion and order criterion via frequency spectrum features without any limitation of the model's phase can be used in reactor noise analysis.The time constant,natural frequency as well as the recovered transfer function of reactors can bhe obtained via the analyzable model based on reactor noise.

  14. Operating Modes Of Chemical Reactors Of Polymerization

    Directory of Open Access Journals (Sweden)

    Meruyert Berdieva

    2012-05-01

    Full Text Available In the work the issues of stable technological modes of operation of main devices of producing polysterol reactors have been researched as well as modes of stable operation of a chemical reactor have been presented, which enables to create optimum mode parameters of polymerization process, to prevent emergency situations of chemical reactor operation in industrial conditions.

  15. Laminar Entrained Flow Reactor (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2014-02-01

    The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

  16. Silica-Immobilized Enzyme Reactors

    Science.gov (United States)

    2007-08-01

    immobilized artificial membrane chromatography and lysophospholipid micellar electrokinetic chromatography . J. Chromatogr. A 1998, 810, 95-103. 50...Journal of Liquid Chromatography and Related Technologies. Air Force Research Laboratory Materials and Manufacturing Directorate Airbase...immobilized enzyme reactors (IMERs) can also be integrated directly to further analytical methods such as liquid chromatography or mass spectrometry.[6] In

  17. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, Warren G.; Basaran, Osman A.; Harris, Michael T.

    1995-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  18. A Simple Tubular Reactor Experiment.

    Science.gov (United States)

    Hudgins, Robert R.; Cayrol, Bertrand

    1981-01-01

    Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

  19. High temperature catalytic membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    1990-03-01

    Current state-of-the-art inorganic oxide membranes offer the potential of being modified to yield catalytic properties. The resulting modules may be configured to simultaneously induce catalytic reactions with product concentration and separation in a single processing step. Processes utilizing such catalytically active membrane reactors have the potential for dramatically increasing yield reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity. Examples of commercial interest include hydrogenation, dehydrogenation, partial and selective oxidation, hydrations, hydrocarbon cracking, olefin metathesis, hydroformylation, and olefin polymerization. A large portion of the most significant reactions fall into the category of high temperature, gas phase chemical and petrochemical processes. Microporous oxide membranes are well suited for these applications. A program is proposed to investigate selected model reactions of commercial interest (i.e. dehydrogenation of ethylbenzene to styrene and dehydrogenation of butane to butadiene) using a high temperature catalytic membrane reactor. Membranes will be developed, reaction dynamics characterized, and production processes developed, culminating in laboratory-scale demonstration of technical and economic feasibility. As a result, the anticipated increased yield per reactor pass economic incentives are envisioned. First, a large decrease in the temperature required to obtain high yield should be possible because of the reduced driving force requirement. Significantly higher conversion per pass implies a reduced recycle ratio, as well as reduced reactor size. Both factors result in reduced capital costs, as well as savings in cost of reactants and energy.

  20. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  1. British high flux beam reactor.

    Science.gov (United States)

    Egelstaff, P A

    1970-10-24

    The neutron scattering technique has become an accepted method for the study of condensed matter. Because of the great scientific and technical value of neutron experiments and the growing body of users, several proposals have been made during the past decade for a nuclear reactor devoted primarily to this technique. This article reviews the reasons for and history behind these proposals.

  2. Thermohydraulic and nuclear modeling of natural fission reactors

    Science.gov (United States)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  3. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  4. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, F-13108 Saint Paul lez Durance (France); Vacelet, H. [CERCA, Romans (France); Dornbusch, D. [Technicatome, Aix en Provence (France)

    2000-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel are discussed. (author)

  5. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1973-07-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  6. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  7. Selection of Type I and Type II Methanotrophic Proteobacteria in a Fluidized Bed Reactor under Non-Sterile Conditions

    Science.gov (United States)

    2011-08-01

    00-00-2011 to 00-00-2011 4. TITLE AND SUBTITLE Selection of Type I and Type II methanotrophic proteobacteria in a fluidized bed reactor under...laboratory- scale fluidized bed reactor was initially inoculated with a Type II Methylocystis-like dominated culture. At elevated levels of dissolved...personal copy Selection of Type I and Type II methanotrophic proteobacteria in a fluidized bed reactor under non-sterile conditions Andrew R. Pfluger a, Wei

  8. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  9. NPR (New Production Reactor) capacity cost evaluation

    Energy Technology Data Exchange (ETDEWEB)

    None

    1988-07-01

    The ORNL Cost Evaluation Technical Support Group (CETSG) has been assigned by DOE-HQ Defense Programs (DP) the task defining, obtaining, and evaluating the capital and life-cycle costs for each of the technology/proponent/site/revenue possibilities envisioned for the New Production Reactor (NPR). The first part of this exercise is largely one of accounting, since all NPR proponents use different accounting methodologies in preparing their costs. In order to address this problem of comparing ''apples and oranges,'' the proponent-provided costs must be partitioned into a framework suitable for all proponents and concepts. If this is done, major cost categories can then be compared between concepts and major cost differences identified. Since the technologies proposed for the NPR and its needed fuel and target support facilities vary considerably in level of technical and operational maturity, considerable care must be taken to evaluate the proponent-derived costs in an equitable manner. The use of cost-risk analysis along with derivation of single point or deterministic estimates allows one to take into account these very real differences in technical and operational maturity. Chapter 2 summarizes the results of this study in tabular and bar graph form. The remaining chapters discuss each generic reactor type as follows: Chapter 3, LWR concepts (SWR and WNP-1); Chapter 4, HWR concepts; Chapter 5, HTGR concept; and Chapter 6, LMR concept. Each of these chapters could be a stand-alone report. 39 refs., 36 figs., 115 tabs.

  10. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  11. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  12. Neutrino Mixing Discriminates Geo-reactor Models

    CERN Document Server

    Dye, S T

    2009-01-01

    Geo-reactor models suggest the existence of natural nuclear reactors at different deep-earth locations with loosely defined output power. Reactor fission products undergo beta decay with the emission of electron antineutrinos, which routinely escape the earth. Neutrino mixing distorts the energy spectrum of the electron antineutrinos. Characteristics of the distorted spectrum observed at the earth's surface could specify the location of a geo-reactor, discriminating the models and facilitating more precise power measurement. The existence of a geo-reactor with known position could enable a precision measurement of the neutrino oscillation parameter delta-mass-squared.

  13. Reactor assessments of advanced bumpy torus configurations

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1983-01-01

    Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of these configurations have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties. Results indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  14. Refurbishment of existing research reactors for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Gessaghi, V. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    Some research reactors have been selected for the development of boron neutron capture therapy (BNCT) in the United States like the Massachusetts Institute of Technology research reactor, the University of Missouri research reactor 2 or the Brookhaven Medical Research Reactor, among others. These reactors have received excellent analyses and designs to accommodate extremely optimized beam shaping assemblies (BSAs) for the proper tuning of neutron spectra and absorption of undesired particles such as photons and fast neutrons. Due to the importance of BNCT in these facilities, the physicists and engineers have used many degrees of freedom for the optimization process.

  15. Sensitivity of reactor multiplication factor to positions of cross-section resonances

    Indian Academy of Sciences (India)

    V GOPALAKRISHNAN; K R VIJAYARAGHAVAN

    2017-09-01

    Neutron–nuclear interaction cross-section is sensitive to neutron kinetic energy and most nuclei exhibit resonance behaviour at specific energies within the resonance energy range, spanning from a fraction of an electron volt to several tens or hundreds of kilo electron volts. The energy positions of these resonances correspond to the excitation energy levels of the compound nucleus that are formed as intermediate states during the interaction. Though these positions, thanks to sophistication in science and technology, are known reasonably precisely for the materials of reactor interest, deviations or spread in this data among different evaluations cannot be ruled out. In this work, the effect of such a spread in the resonance positions of the reactor materials on the multiplication factor of an infinite reactor, is obtained. The study shows that the effect on a thermal reactor is more pronounced than on a fast reactor.

  16. Design of fuzzy PID controller for high temperature pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Badgujar, Kushal D.; Satpute, Satchidanand R.; Revankara, Shripad T.; Lee, John C.; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of)

    2012-10-15

    Control system is most important characteristic to be considered to control spontaneous fission reaction in the design of the nuclear reactor. Recently fuzzy based control systems have been designed and applied as control system for nuclear plants. This article emphasize on controlling the power of the high temperature pebble bed reactor (HTPBR) with the design of Fuzzy proportional integral derivative (PID) controller. A simplified reactor model with point kinetics equation and reactor heat balance equation is used. The reactivity feedback arising from power coefficient of reactivity and Xenon poisoning is also considered. The reactor is operated at various power levels by using fuzzy PID controller. The fuzzy logic eliminates the necessity of the tuning the gains of PID controller each time by extending the finite sets of the PID controller gains.

  17. Optimization and simplification of the concept of non-moderated Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, Elsa; Heuer, Daniel; Allibert, Michel; Doligez, Xavier; Ghetta, Veronique; Le Brun, Christian [LPSC-IN2P3-CNRS/UJF/INPG, LPSC 53 avenue des Martyrs, 38026 Grenoble Cedex (France)

    2008-07-01

    Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria defined by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/{sup 233}U fuel cycle with fluoride salts. The characteristics of the non-moderated TMSR based on a fast neutron spectrum are detailed in this paper: we aimed at designing an optimised TMSR with the simplest configuration. Using a simple kinetic-point model, we analyze the reactor's transient as the total reactivity margins are introduced in the core. We thus confirm, beyond the classical advantages of molten salt reactors, the satisfactory behaviour of the TMSR in terms of safety and the excellent level of stability which can be achieved in such reactors. (authors)

  18. Axial offset as measure of stability of light water nuclear reactor during capacity maneuvering

    Directory of Open Access Journals (Sweden)

    Mark V. Nikolsky

    2015-03-01

    Full Text Available High reliability and security of power unit are required during operation of power unit while maneuvering. They depend on the stability of reactor when transition from one power level to another. The axial offset is a quantitative measure of the reactor stability. It is shown that change of the active core inlet coolant temperature yields an uncontrollable disturbance affecting the axial offset and therefore the reactor stability. To insure the reactor stability the compromise-combined power control method is proposed. Analysis of the influence of temperature of coolant at the magnitude of the axial offset for different regulatory programs is carried out. The change in the depth of immersion of regulators in the active zone for different regulatory programs when the reactor plant daily capacity maneuver is studied.

  19. Small Modular Reactors: Institutional Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Perkowski, Ph.D.

    2012-06-01

    ? Objectives include, among others, a description of the basic development status of “small modular reactors” (SMRs) focused primarily on domestic activity; investigation of the domestic market appeal of modular reactors from the viewpoints of both key energy sector customers and also key stakeholders in the financial community; and consideration of how to proceed further with a pro-active "core group" of stakeholders substantially interested in modular nuclear deployment in order to provide the basis to expedite design/construction activity and regulatory approval. ? Information gathering was via available resources, both published and personal communications with key individual stakeholders; published information is limited to that already in public domain (no confidentiality); viewpoints from interviews are incorporated within. Discussions at both government-hosted and private-hosted SMR meetings are reflected herein. INL itself maintains a neutral view on all issues described. Note: as per prior discussion between INL and CAP, individual and highly knowledgeable senior-level stakeholders provided the bulk of insights herein, and the results of those interviews are the main source of the observations of this report. ? Attachment A is the list of individual stakeholders consulted to date, including some who provided significant earlier assessments of SMR institutional feasibility. ? Attachments B, C, and D are included to provide substantial context on the international status of SMR development; they are not intended to be comprehensive and are individualized due to the separate nature of the source materials. Attachment E is a summary of the DOE requirements for winning teams regarding the current SMR solicitation. Attachment F deserves separate consideration due to the relative maturity of the SMART SMR program underway in Korea. Attachment G provides illustrative SMR design features and is intended for background. Attachment H is included for overview

  20. Savannah River Site production reactor technical specifications. K Production Reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

  1. CFD Simulation on Ethylene Furnace Reactor Tubes

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Different mathematical models for ethylene furnace reactor tubes were reviewed. On the basis of these models a new mathematical simulation approach for reactor tubes based on computational fluid dynamics (CFD) technique was presented. This approach took the flow, heat transfer, mass transfer and thermal cracking reactions in the reactor tubes into consideration. The coupled reactor model was solved with the SIMPLE algorithm. Some detailed information about the flow field, temperature field and concentration distribution in the reactor tubes was obtained, revealing the basic characteristics of the hydrodynamic phenomena and reaction behavior in the reactor tubes. The CFD approach provides the necessary information for conclusive decisions regarding the production optimization, the design and improvement of reactor tubes, and the new techniques implementation.

  2. Optimized Design and Discussion on Middle and Large CANDLE Reactors

    Directory of Open Access Journals (Sweden)

    Xiaoming Chai

    2012-08-01

    Full Text Available CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scale fission nuclear energy utilization for thousands of years, benefitting the whole of society. For many developing countries with a large population and high energy demands, such as China and India, middle (1000 MWth and large (2000 MWth CANDLE fast reactors are obviously more suitable than small reactors [2]. In this paper, the middle and large CANDLE reactors are investigated with U-Pu and combined ThU-UPu fuel cycles, aiming to utilize the abundant thorium resources and optimize the radial power distribution. To achieve these design purposes, the present designs were utilized, simply dividing the core into two fuel regions in the radial direction. The less active fuel, such as thorium or natural uranium, was loaded in the inner core region and the fuel with low-level enrichment, e.g. 2.0% enriched uranium, was loaded in the outer core region. By this simple core configuration and fuel setting, rather than using a complicated method, we can obtain the desired middle and large CANDLE fast cores with reasonable core geometry and thermal hydraulic parameters that perform safely and economically; as is to be expected from CANDLE. To assist in understanding the CANDLE reactor’s attributes, analysis and discussion of the calculation results achieved are provided.

  3. Actinides in irradiated graphite of RBMK-1500 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Plukienė, R., E-mail: rita@ar.fi.lt; Plukis, A.; Barkauskas, V.; Gudelis, A.; Gvozdaitė, R.; Duškesas, G.; Remeikis, V.

    2014-10-01

    Highlights: • Activation of actinides in the graphite of the RBMK-1500 reactor was analyzed. • Numerical modeling using SCALE 6.1 and MCNPX was used for actinide calculation. • Measurements of the irradiated graphite sample were used for model validation. • Results are important for further decommissioning process of the RBMK type reactors. - Abstract: The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK

  4. Multimegawatt space power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dearien, J.A.; Whitbeck, J.F.

    1989-01-01

    In response to the need of the Strategic Defense Initiative (SDI) and long range space exploration and extra-terrestrial basing by the National Air and Space Administration (NASA), concepts for nuclear power systems in the multi-megawatt levels are being designed and evaluated. The requirements for these power systems are being driven primarily by the need to minimize weight and maximize safety and reliability. This paper will discuss the present requirements for space based advanced power systems, technological issues associated with the development of these advanced nuclear power systems, and some of the concepts proposed for generating large amounts of power in space. 31 figs.

  5. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  6. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  7. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  8. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser

    2012-01-01

    Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.

  9. A reduced fidelity model for the rotary chemical looping combustion reactor

    KAUST Repository

    Iloeje, Chukwunwike O.

    2017-01-11

    The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel fraction for the selected oxygen carrier materials.

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

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

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

  11. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  12. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  13. Status of reactor shielding research in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties.

  14. Gas-liquid autoxidation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morbidelli, M.; Paludetto, R.; Carra, S.

    1986-01-01

    A procedure for the simulation of autoxidation gas-liquid reactors has been developed based both on mathematical models and laboratory experiments. It has been shown that the complex radical chain mechanism of the autoxidation process can be simulated through two global parallel reactions, whose rates are obtained by assuming pseudo-steady-state concentration values for all the radical species involved. Using ethylbenzene autoxidation as a model reaction, an experimental analysis has been performed in order to estimate all the kinetic parameters of the model. The effect of the interaction between gas-liquid mass-transfer phenomena and the complex kinetic mechanism on the overall performance of an autoxidation reactor has been examined in detail within the framework of the liquid film model.

  15. Transport simulation for EBT reactor

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, T.; Uckan, N.A.; Jaeger, E.F.

    1983-08-01

    Transport simulation and modeling studies for the ELMO Bumpy Torus (EBT) reactor are carried out by using zero-dimensional (0-D) and one-and-one-half-dimensional (1 1/2-D) transport calculations. The time-dependent 0-D model is used for global analysis, whereas the 1 1/2-D radial transport code is used for accurate determination of density, temperature, and ambipolar potential profiles and of the role of these profiles in reactor plasma performance. Analysis with the 1 1/2-D transport code shows that profile effects near the outer edge of the hot electron ring lead to enhanced confinement by at least a factor of 2 to 5 beyond the simple scaling that is obtained from the global analysis. The radial profiles of core plasma density and temperatures (or core pressure) obtained from 1 1/2-D transport calculations are found to be similar to those theoretically required for stability.

  16. Nuclear reactor alignment plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  17. Fluidized bed coal combustion reactor

    Science.gov (United States)

    Moynihan, P. I.; Young, D. L. (Inventor)

    1981-01-01

    A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

  18. Fast breeder reactor protection system

    Science.gov (United States)

    van Erp, J.B.

    1973-10-01

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

  19. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  20. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  1. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  2. Transient behavior of a nuclear reactor coupled to an accelerator

    Science.gov (United States)

    Sadineni, Suresh Babu

    Accelerator Driven Systems (ADS) present one of the most viable solutions for transmutation and effective utilization of nuclear fuel. Spent fuel from reactors will be partitioned to separate plutonium and other minor actinides to be transmuted in the ADS. Without the ADS, minor actinides must be stored at a geologic repository for long periods of time. One problem with ADS is understanding the control issues that arise when coupling an accelerator to a reactor. "ADSTRANS" was developed to predict the transient behavior of a nuclear reactor coupled to an accelerator. It was based on MCNPX, a radiation transport code developed at the LANL, and upon a numerical model of the neutron transport equation. MCNPX was used to generate the neutron "source" term that occurs when the accelerator is fired. ADSTRANS coupled MCNPX to a separate finite difference code that solved the transient neutron transport equation. A cylindrical axisymmetric reactor with steel shielding was considered for this analysis. Multiple neutron energy groups, neutron precursor groups and neutron poisons were considered. ENDF/B cross-section data obtained through MCNPX was also employed. The reactor was assumed to be isothermal and near zero power level. Unique features of this code are: (1) it predicts the neutron behavior of an ADS for different reactor geometry, material concentration, both electron and proton particle accelerators, and target material, (2) it develops input files for MCNPX to simulate neutron production, runs MCNPX, and retrieves information from the MCNPX output files. Neutron production predicted by MCNPX for a 20 MeV electron accelerator and lead target was compared with experimental data from the Idaho Accelerator Center and found to be in good agreement. The spatial neutron flux distribution and transient neutron flux in the reactor as predicted by the code were compared with analytical solutions and found to be in good agreement. Fuel burnup and poison buildup were also as

  3. Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne L.

    2017-08-23

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  4. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  5. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J.H.

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  6. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  7. Investigation of molten salt fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kubota, Kenichi; Konomura, Mamoru [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2002-05-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  8. DNA-Based Enzyme Reactors and Systems

    Directory of Open Access Journals (Sweden)

    Veikko Linko

    2016-07-01

    Full Text Available During recent years, the possibility to create custom biocompatible nanoshapes using DNA as a building material has rapidly emerged. Further, these rationally designed DNA structures could be exploited in positioning pivotal molecules, such as enzymes, with nanometer-level precision. This feature could be used in the fabrication of artificial biochemical machinery that is able to mimic the complex reactions found in living cells. Currently, DNA-enzyme hybrids can be used to control (multi-enzyme cascade reactions and to regulate the enzyme functions and the reaction pathways. Moreover, sophisticated DNA structures can be utilized in encapsulating active enzymes and delivering the molecular cargo into cells. In this review, we focus on the latest enzyme systems based on novel DNA nanostructures: enzyme reactors, regulatory devices and carriers that can find uses in various biotechnological and nanomedical applications.

  9. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  10. Radioactivity Level in the Environment around Xi’an Pulsed Reactor in 2006-2010%2006-2010年西安脉冲堆周边环境介质放射性水平

    Institute of Scientific and Technical Information of China (English)

    李华; 于青玉; 唐秀欢; 张继红; 马燕; 宋晓靓

    2014-01-01

    介绍了2006—2010年西安脉冲堆堆址周边气溶胶、土壤、水、代表性生物等环境介质放射性水平的监测调查工作。调查结果显示:2006—2010年,西安脉冲堆周边环境介质中,气溶胶总α、总β活度浓度平均值为0.1、0.6 Bq/m3;地表水总α、总β和137 Cs平均活度浓度分别为101.4、249.8、2.4 mBq/L;地下水总α、总β和137 Cs平均活度浓度分别为46.7、90.4、2.1 mBq/L;代表性生物中总α、总β和137 Cs平均活度浓度分别为0.6、58.2、2.2×10-2 Bq/kg (鲜重);土壤中总α、总β、137 Cs、226 Ra、232 Th、40 K平均活度浓度分别为386、793、3.0、36、48、548 Bq/kg (干重)。%The results of radioactivity monitoring of environmental media such as air ,water ,representative bi-omaterial ,and soil during 2006 to 2010 were presented .The average concentrations of totalα,total βin the air aerosol were 0 .1 ,0 .6 Bq/m3 .Totalα,totalβand 137 Cs in surface water averaged 101 .4 ,249 .8 ,2 .4 mBq/L ,while corresponding values in underground water and biomaterial were 46 .7 ,90 .4 ,2 .1 mBq/L ,and 0 .6 ,58 .2 ,2 .2 × 10-2 Bq/kg(wet) ,respectively .In soil ,the average levels of total α,total β,137 Cs , 226 Ra ,232 Th and 40 K were 386 ,793 ,3 .0 ,36 ,48 ,548 Bq/kg (dry ) ,respectively .

  11. Alcohol synthesis in a high-temperature slurry reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, G.W.; Marquez, M.A.; McCutchen, M.S. [North Carolina State Univ., Raleigh, NC (United States)

    1995-12-31

    The overall objective of this contract is to develop improved process and catalyst technology for producing higher alcohols from synthesis gas or its derivatives. Recent research has been focused on developing a slurry reactor that can operate at temperatures up to about 400{degrees}C and on evaluating the so-called {open_quotes}high pressure{close_quotes} methanol synthesis catalyst using this reactor. A laboratory stirred autoclave reactor has been developed that is capable of operating at temperatures up to 400{degrees}C and pressures of at least 170 atm. The overhead system on the reactor is designed so that the temperature of the gas leaving the system can be closely controlled. An external liquid-level detector is installed on the gas/liquid separator and a pump is used to return condensed slurry liquid from the separator to the reactor. In order to ensure that gas/liquid mass transfer does not influence the observed reaction rate, it was necessary to feed the synthesis gas below the level of the agitator. The performance of a commercial {open_quotes}high pressure {close_quotes} methanol synthesis catalyst, the so-called {open_quotes}zinc chromite{close_quotes} catalyst, has been characterized over a range of temperature from 275 to 400{degrees}C, a range of pressure from 70 to 170 atm., a range of H{sub 2}/CO ratios from 0.5 to 2.0 and a range of space velocities from 2500 to 10,000 sL/kg.(catalyst),hr. Towards the lower end of the temperature range, methanol was the only significant product.

  12. A gas-cooled reactor surface power system

    Science.gov (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  13. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    Science.gov (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.

    2015-02-01

    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  14. OECD/NEA STUDY ON THE ECONOMICS AND MARKET OF SMALL REACTORS

    Directory of Open Access Journals (Sweden)

    ALEXEY LOKHOV

    2013-11-01

    Full Text Available According to the OECD/NEA estimates, nuclear power plants (NPPs, whether with a large reactor or with small modular reactors (SMRs, are competitive with many other electricity generation technologies in a significant number of cases, one of the exceptions being natural gas in the USA with the current level of prices. However, SMRs have particular features and requirements setting conditions for their deployment. This paper presents the preliminary analysis by OECD/NEA of the economics, opportunities, and market for small nuclear reactors.

  15. New options for developing of nuclear energy using an accelerator-driven reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroshi

    1997-09-01

    Fissile fuel can be produced at a high rate using an accelerator-driven Pu-fueled subcritical fast reactor. Thus, the necessity of early introduction of the fast reactor can be moderated. High reliability of the proton accelerator, which is essential to implementing an accelerator-driven reactor in the nuclear energy field can be achieved by a slight extension of the accelerator`s length, with only a small economical penalty. Subcritical operation provides flexible nuclear energy options including high neutron economy producing the fuel, transmuting high-level wastes, such as minor actinides, and of converting efficiently the excess Pu and military Pu into proliferation-resistant fuel.

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

    Science.gov (United States)

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

    2017-01-01

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

  17. [Study on hydrogen autotrophic denitrification of bio-ceramic reactor].

    Science.gov (United States)

    Chen, Dan; Wang, Hong-Yu; Song, Min; Yang, Kai; Liu, Chen

    2013-10-01

    Nitrate wastewater is processed in a bio-ceramic reactor based on hydrogen autotrophic denitrification. The implementation procedure of biological denitrification by hydrogen autotrophic denitrification was investigated. The effects of hydraulic retention time, influent nitrate load, influent pH, temperature and the amount of hydrogen were assessed throughout this trial. The results showed that the removal rate of NO-(3) -N was 94. 54% and 97. 47% when the hydraulic retention time was 24 h and 48 h, respectively. When the hydraulic retention time was in the range of 5-16 h, the removal rate gradually dropped with the shortening of the hydraulic retention time. When the influent NO-(3) -N concentration was low, with the increase in the influent NO-(3) -N concentration, the degradation rate also increased. The denitrification was inhibited when the NO-(3) -N concentration was higher than 110 mg.L-1. Neutral and alkaline environment was more suitable for the reactor. The reactor showed a wide range of temperature adaptation and the optimum temperature of the reactor was from 25 to 30 degrees C. When hydrogen was in short supply, the effect of denitrification was significantly reduced. These results indicated the specificity of hydrogen utilization by the denitrifying bacteria. The effluent nitrite nitrogen concentration was maintained at low levels during the operation.

  18. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

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

    Directory of Open Access Journals (Sweden)

    Alba Arratibel Plazaola

    2017-01-01

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

  20. Application of the MACCS code to DOE production reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    O' Kula, K.R.; East, J.M. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1991-01-01

    A three-level probabilistic risk assessment (PRA) of the special materials production reactor operation at the US Department of Energy's (DOE's) Savannah River site (SRS) has been completed. The goals of this analysis were to: (1) analyze existing margins of safety provided by the heavy water reactor (HWR) design challenged by postulated severe accidents; (2) compare measures of risk to the general public and on-site workers to guideline values, as well as to those posed by commercial reactor operation; and (3) develop the methodology and data base necessary to determine the equipment, human actions, and engineering systems that contribute significantly to ensuring overall plant safety. In particular, the third point provides the most tangible benefit of a PRA since the process yields a prioritized approach to increasing safety through design and operating practices. This paper describes key aspects of the consequence analysis portion of the SRS PRA: Given the radiological releases quantified through the level-2 PRA analysis, the consequences to the off-site general public and to the on-site SRS workforce are calculated. This analysis, the third level of the PRA, is conducted primarily with the MACCS 1.5 code. The level-3 PRA yields a probabilistic assessment of health and economic effects based on meteorological conditions sampled from site-specific data.